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

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| issue date = 02/03/1986
| issue date = 02/03/1986
| title = Failure of Lifting Rig Attachment, While Lifting Upper Guide Structure at St. Lucie Unit 1
| title = Failure of Lifting Rig Attachment, While Lifting Upper Guide Structure at St. Lucie Unit 1
| author name = Jordan E L
| author name = Jordan E
| author affiliation = NRC/IE
| author affiliation = NRC/IE
| addressee name =  
| addressee name =  
Line 14: Line 14:
| page count = 7
| page count = 7
}}
}}
{{#Wiki_filter:SSINS No.: 6835 IN 86-06 UNITED STATES NUCLEAR REGULATORY
{{#Wiki_filter:SSINS No.: 6835 IN 86-06 UNITED STATES


COMMISSION
NUCLEAR REGULATORY COMMISSION


OFFICE OF INSPECTION
OFFICE OF INSPECTION AND ENFORCEMENT


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


LIFTING THE UPPER GUIDE STRUCTURE
ST. LUCIE UNIT 1
 
AT ST. LUCIE UNIT 1  


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


permit (CP). .
permit (CP).                                                       .


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


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


significant
expected that recipients will review this information for applicability to their


problem that occurred during the movement of a heavy load over the reactor core. It is expected that recipients
facilities and consider actions, if appropriate, to preclude similar problems at


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


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


to their facilities
is required.


and consider actions, if appropriate, to preclude similar problems at their facilities.
==Description of Circumstances==
 
:
However, the suggestions
On November 6, 1985, while lifting the upper guide structure from the St.
 
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
Lucie Unit 1 reactor vessel, licensee personnel noticed the lifting rig tilt.


and moved to its normal parking position in the refueling
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


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:
three attachment points. An attempt was promptly made to lower the'load back
The upper guide structure


is shown in Figure 1. It is supported
to its installed position, but the load cells indicated binding, so the attempt


in the reactor vessel by it§..upper
was terminated after lowering the load a few inches. The 50-ton load was left


flange. It is aligned by eight alignment
suspended about 8 feet above the reactor core.


keys, four at the top and fourat the bottom. The structure
The licensee declared an Unusual Event. Core alteration containment integrity


fits down inside the core support barrel, just'above
was enhanced by resuming full use of the airlocks. Temporary primary manway


the fuel assemblies (see Figure 2). The fuel assembly alignment
covers were installed on bcth hot and cold legs to enhance the nozzle dams.


plate is the bottom component
Survey transits were set up and procedures implemented to monitor the rig and


of the upper guide structure.
load for any shifts in their positions.


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


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


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


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


procedure
refueling pool.


for attaching
8601290056


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


and torquing each bolt to 50 ft-lbs. The licensee's
the two guide pins that align the rig with the reactor vessel.


procedure omitted the step concerning
Discussion:
The upper guide structure is shown in Figure 1. It is supported in the reactor


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


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


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


of the bolt that had pulled loose indicated
alignment plate is the bottom component of the upper guide structure.


that part of the last thread was stripped.
The lifting rig is attached to the upper guide structure by three vertically


It is assumed that this bolt cross-threaded
oriented bolts. These bolts are attached from above the water line by torque


or bound due to rig to guide structure
tools that run down the hollow columns of the rig (see Figure 3). Combustion


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


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


and reached the 50 ft-lb torque requirement
omitted the step concerning the check for thread engagement. Subsequent


with only part of one thread engaged. During the lift, the few inches of unengaged
inspection of the bolt that had pulled loose indicated that part of the last


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


in an imperceptible
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


tilt. The resulting
lift, the few inches of unengaged bolt shaft were pulled through the lifting


lateral load was initially
rig until the bolt head rested on the rig's surface at the bottom of the


s6pported
column, resulting in an imperceptible tilt. The resulting lateral load was


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


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


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


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


engaged bolt to slip free. This caused the observed motion and tilt.After the guide structure
free. This caused the observed motion and tilt.


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


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


lifting rig and leveled, it was moved to its normal parking position in the refueling
The short attachment bolts and torque tools were then replaced with full-length


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


bolts and torque tools were then replaced with full-length
of the three hollow columns of the lifting rig. This has the advantage of


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
making anyflmck of full thread engagement more apparent to the personnel


of making anyflmck of full thread engagement
attaching the rig. The guide structure was subsequently returned to its


more apparent to the personnel attaching
installed position using the long attachment bolts.


the rig. The guide structure
The licensee has not yet decided whether to permanently modify the attachment


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


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


position using the long attachment
measuring thread engagement.


bolts.The licensee has not yet decided whether to permanently
The potential consequences of dropping heavy loads into the open reactor vessel


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


bolts. The licensee plans to review all reactor-related
Near Spent Fuel." The concern for a UGS drop is that fuel assemblies might be


lifts for adequacy of the procedures
sufficiently damaged to release the radioactive gases and iodines held within


to ensure proper lift rig attachment, including
the fuel-clad gap. Under the reduced containment integrity requirements for


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


for measuring
dose limits of 10 CFR 100 to be exceeded.


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


The potential
the NRC determined that further calculations were not required after reviewing


consequences
initial calculations previously submitted by other reactor facilities in


of dropping heavy loads into the open reactor vessel were addressed
response to Phase II of USI A-36. Some indication of the consequences of a


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


Safety Issue (USI) A-36, "Control of Heavy Loads Near Spent Fuel." The concern for a UGS drop is that fuel assemblies
Engineering for a reactor vessel head drop at Waterford 3. The head drop


might be sufficiently
calculations assumed the reactor vessel head was sufficiently tilted at impact


damaged to release the radioactive
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


gases and iodines held within the fuel-clad
second, and the resulting vertical stresses imposed on the fuel were not


gap. Under the reduced containment
sufficient to rupture the cladding.


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


requirements
would have been substantially less than the UGS response velocity calculated


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


mode, damage to several fuel assemblies
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


might cause the radiation dose limits of 10 CFR 100 to be exceeded.
scenario. If substantial misalignment occurred, the fuel could be subjected to


IN 86-06 February 3, 1986 Plant specific calculations
additional axial loading. Significant misalignment could not occur without


were not made for a UGS drop at St. Lucie because the NRC determined
substantial impact damage to the eight keys and keyways, which would also


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


were not required after reviewing initial calculations
this basis, significant radioactive gas release is considered to be unlikely, although it has not been shown to be impossible.


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


submitted
have any questions regarding this matter, please contact the Regional


by other reactor facilities
Administrator of the appropriate NRC regional office or this office.


in response to Phase II of USI A-36. Some indication
ar    Jordan, Director


of the consequences
Divisio of Emergency Preparedness


of a UGS drop at St. Lucie can be gained from calculations
and  gineering Response


performed
Office of Inspection and Enforcement


by Combustion
Technical Contacts:  S. M. Long, IE


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


for a reactor vessel head drop at Waterford
(301) 492-9735 Attachments:
1. Figure 1, Upper Guide Structure Assembly


3. The head drop calculations
2. Figure 2, Reactor Internals Assembly


assumed the reactor vessel head was sufficiently
3    Figure 3, Upper Guide Structure Lifting Rig


tilted at impact to directly strike the UGS with the UGS at rest in its normal installed
4. List of Recently Issued IE Information Notices


posi-tion. The calculated
Attachment 1 IN 86-06 February 3, 1986 EXPANSION


response velocity of the Waterford
COMPENSATING


UGS was 28 feet per second, and the resulting
RING


vertical stresses imposed on the fuel were not sufficient
CEA


to rupture the cladding.If the St. Lucie UGS had dropped from an 8 foot elevation, its striking velocity would have been substantially
SHROUD


less than the UGS response velocity calculated
GRID


for the Waterford
ASSEMBLY.


head drop. However, the potential
CEA SHROUDS


for misalignment
FUEL ASSEMBLY


of the recesses in the bottom of the UGS (i.e., the fuel assembly alignment
ALIGNMENT PLATE


plate)with the fuel assembly upper end fitting posts was not addressed
Figure 1: Upper Guide Structure Assembly


by the Waterford scenario.
(St. Lucie Unit 1 FSAR-figure 4.2-10)


If substantial
UPPER GUIDE


misalignment
-STRUCTURE SUPPORT


occurred, the fuel could be subjected
PLATE


to additional
Attachment 2 IN 86-06 February 3, 1986 CEA


axial loading. Significant
SHROUD


misalignment
IN-CORE


could not occur without substantial
-INSTRUMENTATION


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
OUTLET                                                  GUIDE TUBE


radioactive
CORE


gas release is considered
' SUPPORT


to be unlikely, although it has not been shown to be impossible.
BARREL


No specific action or written response is required by this notice. If you have any questions
FUEL


regarding
ALIGNMENT


this matter, please contact the Regional Administrator
PINS


of the appropriate
CORE


NRC regional office or this office.ar Jordan, Director Divisio of Emergency
-SUPPORT


===Preparedness===
ASSEMBLY
and gineering


Response Office of Inspection
Am. 3-7/85 Figure 2: Reactor Internals Assembly


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


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


Contacts:
CLEVIS ASSEMBLY
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
WORKING PLATFORM


Assembly 3 Figure 3, Upper Guide Structure
I NSTRUMENT


Lifting Rig 4. List of Recently Issued IE Information
STALK OPENING


Notices
- COLUMN


Attachment
GUIDE BUSHING FOR


1 IN 86-06 February 3, 1986 EXPANSION COMPENSATING
R.V. GUIDE PIN


RING CEA SHROUD GRID ASSEMBLY.CEA SHROUDS FUEL ASSEMBLY ALIGNMENT
a PLACES)
                                                          -  00
            LIFT BOLT


PLATE Figure 1: Upper Guide Structure
Fi gure 3: Upper Guide Structure Lifting Rig


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


4.2-10)
Attachment 4 IN 86-06 February 3, 1986 LIST OF RECENTLY ISSUED
UPPER GUIDE-STRUCTURE


SUPPORT PLATE Attachment
IE INFORMATION NOTICES


2 IN 86-06 February 3, 1986 CEA SHROUD IN-CORE-INSTRUMENTATION
Tnfn,'mnfi n


GUIDE TUBE OUTLET CORE' SUPPORT BARREL FUEL ALIGNMENT PINS CORE-SUPPORT ASSEMBLY Am. 3-7/85 Figure 2: Reactor Internals
===Notice No.     Subject===
                                              Date oT


Assembly (St. Lucie Unit 1 FSAR figure 4.2-7)
Issue        T-Issud tn
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)
Issu----    ud -- t
Attachment


4 IN 86-06 February 3, 1986 LIST OF RECENTLY ISSUED IE INFORMATION
86-05          Main Steam Safety Valve Test  1/31/86      All PWR facilities


NOTICES T nfn,'mnfi
Failures And Ring Setting                  holding an OL or


n Notice No. Subject Date oT Issue T-Issud tn Issu---- --ud t 86-05 86-04 86-03 86-02 86-01 85-101 85-100 85-99 85-98 Main Steam Safety Valve Test 1/31/86 Failures And Ring Setting Adjustments
Adjustments                                 CP


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


Due To Loss Of 1/31/86 Power To Integrated
Power To Integrated Control                facilities holding


Control System At A Pressurized
System At A Pressurized Water              an OL or CP


Water Reactor Designed By Babcock& Wilcox Potential
Reactor Designed By Babcock


Deficiencies
& Wilcox


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


===Qualification===
Environmental Qualification                 facilities holding
Of Limitorque


Motor Valve Operator Wiring Failure Of Valve Operator 1/6/86 Motor During Environmental
Of Limitorque Motor Valve                   an OL or CP


Qualification
Operator Wiring


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


1/6/86 Check Valve Causes Loss Of Feedwater
Motor During Environmental                  facilities holding


System Integrity And Water-Hammer
Qualification Testing                      an OL or CP


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


of 10 CFR 21 12/31/85 To Consulting
Check Valve Causes Loss Of                  facilities holding


Firms Providing Training Rosemount
Feedwater System Integrity                  an OL or CP


Differential
And Water-Hammer Damage


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


Zero Point Shift Cracking In Boiling-Water-
To Consulting Firms Providing              facilities holding
12/31/85 Reactor Mark I And Mark II Containments


Caused By Failure Of The Inerting System All PWR facilities
Training                                    an OL or CP


holding an OL or CP All power reactor facilities
85-100        Rosemount Differential        12/31/85    All power reactor


holding an OL or CP All power reactor facilities
Pressure Transmitter Zero                  facilities holding


holding an OL or CP All power reactor facilities
Point Shift                                an OL or CP


holding an OL or CP All power reactor facilities
85-99          Cracking In Boiling-Water-    12/31/85    All BWR facilities


holding an OL or CP All power reactor facilities
Reactor Mark I And Mark II                  having a Mark I or


holding an OL or CP All power reactor facilities
Containments Caused By Failure              Mark II containment


holding an OL or CP All BWR facilities
Of The Inerting System


having a Mark I or Mark II containment
85-98          Missing Jumpers From Westing- 12/26/85      All Westinghouse


===All Westinghouse===
house Reactor Protection                    designed PWR
designed PWR facilities


holding an OL or CP Missing Jumpers From Westing- 12/26/85 house Reactor Protection
System Cards For The Over-                  facilities holding


System Cards For The Over-Power Delta Temperature
Power Delta Temperature Trip                an OL or CP


Trip Function OL = Operating
Function


License CP = Construction
OL = Operating License


Permit}}
CP = Construction Permit}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Latest revision as of 03:07, 24 November 2019

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)
v • d • e


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


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