ML20086Q110
| ML20086Q110 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 12/20/1991 |
| From: | Mcmeekin T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 91-23, NUDOCS 9112300001 | |
| Download: ML20086Q110 (25) | |
Text
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_o j Dukeikuer Cstnpany.
.- T C Ma: tun McGuire Nuclear Generakon department Vice />rsident
-!!?00 Hagers Ferry ReadinfG01A)
(iO4)Eih480('
- fluttersville MC2Roi8 89%
(104)3754809fn DUKE POWER December 20, 1991 U.S. Nuclear Regulatory Commission Dccument Control Desk Washington, D.C.
20555 subject:- McGuire' Nuclear Station Unit 1~
Docket No. 50-369 Voluntary SEsecial-Report i
entlemen:
Attathed is Voluntary Special - Report concerning. an -incident in
- which the reactor vessel -lower internals contacted the reactor-H vossal during removal.
This D event is considered to be of no
,l significance with respect to the health and safety of the public.
l l
feryLtruly yours, 0
I.
hk T. C. McMeekin NGA/cbl-Attachment-xc:
Mr. S. D.: Ebneter Administrator, Region II-
'U.S. Nuclear' Regulatory Commission 101 Marietta-St., NW, Suite-2900.
Atlanta, GA 30323-INPO Records Center Suite-1500
=1100 circle 75 Parkway
- Atlanta,~GA= 30339*
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Mr. Tim Reed"
'U.S. Nuc lear Regulatory: Commission Officetof Nuclear. Reactor: Regulation Washington, D.C.
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NRC-. Resident. Inspector
-McGuire: Nuclear Statinn aa01.TL
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McGUIRE SAFETY REVIEW GROUP INPLANT REVIEW IEPORT 1.
REPORT NUMBER: 91-23 2.
DATE OF REVIEW: October 8, 1991 - December 16, 1991 3.
SUBJECT DESCRIPTION: A review was conducted of the circumstances related to the incident described in Problem Investi.gation Report (PIR) 1-M91-0177.
The purpose of the review was to determine the causes of the incident and to identify corrective actions to preclude the recurrence of the incident and associated problems.
4.
EVALUATION AND COMKENT: PIR l-M91-0177 documented an incident in which the reactor vessel lower internals contacted the reactor vessel during removal.
Later, during the move to place the lower internals on its storage stand in tre deep end of the refueling canal, the seal surface protective ring and the lifting rig guide bushing contacted the refueling canal liner plate.
4,1
Background
The reactor vessel lower internals consist of the follcwing components: core barrel, core baffle, lower core plata and support colunes, neutron shield pads, core support plate, and secondary cort support assembly (See page 18).
The total weight of the lower internals and lifting rig is approximately 208,000 pounds and the total length is approximately 400 inches.
The lower internale were being removed for the 10 year Inser" ice Inspection (ISI) on the reactor vessel and associated equipment, and for implementation of the internals upflow modification.
4.2 Deecription of Event Prior to the incident, the upper interna ls had been removed for defueling and were stored in the deep ead of the refueling canal.
All fuel had been removed and was storod in the Spent Fuel Pool, which had been isolated from the refueling cavity by closing valve 1KF-122, Fuel Transfer Tube Isolation Valve. Extensive planning and training were performed for *he removal of the lower
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4 DPC/MNS INPIANT REVIEW No. 91-23 PAGE 2 internals.
It is estimated that over 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> were expended in preparation for the lift.
Four r'.ays of training for the support crew and Technical Support personnel was conducted in August.
Training for the crane personnel was conducted one on one prior to the lift.
A comprehensive prejob briefing was conducted with all personnel involved in the lift. The briefing covered the followings Step by step procedure walk through Idertification of the person (s) responsible for each step Detailed discus sion of critical steps (gauge installation and verification, EMF disabling)
Communication requirements Camera locations and video requirements Actions to be taken in the event communications or video were lost Radiation Protection concerns.
On October 4, 1991, preparations for removal of the lower internals began.
Personnel directly involved in the work were a Technical Support representative, two Polar Crane Operators, a Flagman, two Radiation Protection technicians and a Radiation Protection Supervisor. All prerequisites required for the activity by procedure MP/0/A/7150/01, Reactor' Vessel Lower Internals Removal and Installation, were satisfactorily completed.
The lift occurred on October 8, 1991. Jk dry run moving the lif t rig horizontally was performed as required by procedure to verify Polar Crane match marks at the lower internals storage stand and reactor vessel and to check for interferences with the Manipulator Crane. The Control Room was-notified that the lower internals lift was to commence.
At this point in the evolution, the Crane Operators were receiving instructions from a Flagman who was pr.ysically positioned on the Operating Deck.
The internals were raised to the height necessary for the lower internals elevation gauge to be installed.
(The seal surf ace protective ring was stored on the lif tir.g rig.
The reactor head crew had been unable to remove the seal surfaca
e DPC/HNS INPLANT REVIEW No. 91-23 PAGE 3 protective ring from the lifting rig earlier at the upper internals storage stand because the rods used to disengage the ring were several inches too short. The decision was made by Technical Support personnel to conduct the lift with the protective ring in place.
The seal surface protective ring could have been removed fr-t.ie lift rig and installed on the reactor vessel flange.) The deal surface protective ring interfered with the installation of the elevatior. gauge.
The gauge was instelled at the correct location, but it was not at the. correct elevation.-
After the gauge was installed, all personnel except the Crane Operators were required to leave upper containment. Radiation Protection retreated to the personnel air lock.
These actions were taken because of an expected increase in dcse rates associated with the lift.
(The general area dose rate prior to the lift was 1 to 2 mR/hr.
The highest dose rate measured during the lift was 1060 mR/hr.)
Communication with the Crane Operator and control of the evolution was turned over from the Flagman to Technical Support personnel located outside upper containment at the Video Control Station.
Tt.s lift proceeded (See page 19).
When the lower internals flange broke the water surface, radiation monitor lEMF16 on the refueling crane bridge alarmed, activating the centaAnment evacuation alarm.
This alarm remained activated throughout the incident. A previous procedure step required 1 EMF 16 to be disabled; therefore, i
personnel involved in the lift did not expect it to alarm.
Instrument and Electrical personnel had pulled a relay to disable 1 EMF 16 but this action did not silence the alarm in the Reactor Building.
(The relay that wac removed silenced the alarm in the control Room only.) There was no specific guidt.nce (procedure or work request) addressing how to disable the EMF.
The prejob briefing had specifically addressed that 1EHF16 would be disabled.
The Crane operators were surprised by the alarm, but believed that it would be resolved by Instrument and Electrica) personnel on standby for the lift.
In addition, they were wearing transmitting dosimeters (Merlin Gerin) and knew that Radiation Protection l
personnel were continuously monitoring the dose rates by direct readout outside upper containment.
In response to the alarm, the Control Room was notified by Radiation Protection personnel that Radiation Protection was monitoring the radiological conditions.
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DPC/MNS INPLANT REVIEW No. 91-03 PAGE 4 The Technicsl Support person was controlling the lift from the Video Control Station outside of upper containment in order to reduce dose.
There were five video monitors available. One monitor provided an overall view of upper wontainment, one monitor provided a view of the elevation gauge and three monitors provided views of the keyways on the lower internals stand in the deep end of the refueling canal. One of the three underwater cameras being used to provide a view of a storage stand keyway was rotated so that it provided a view of the vessel flange.
This view was rotated 90 degrees on the screen.
The Technical Support person was utilizing two of five monitors for the lift and horizontal move of the lower internals.
One monitor provided a view of the elevation gauge and the other monitor provided a view of the vessel flange area.
Once the Technical Support person thought he saw the " shock absorbers" (secondary core support assembly) clear the vessel flange, he instructed the Crane Operator to make the horizontal move to the deep end of the canal.
The Crane Operator began the horizontal move at high trolley speed as directed by the procedure and bumped the vessel. The Crane Operator relea ed the trolley and allowed the crane to coast back to neutral.
He lifted some more and began to move: aorizontally a second time at L slower speed, bumped the vessel a second time and stopped the horizontal movement again. He lifted again. During the third horizontal movement, he continued lifting until the lower internals cleared the vessel flange and then moved the lower internals to the deep end of the refueling canal.
At the deep end, the Crane Operator lowered the lower internals and then bridged the crane to centerline of the Storage Stand.
The Crane Operator began to lower the internals further. The Crane Operator was unable to see the Storage Stand 0 degree position markers because of the crane configuration. The protective ring and the lifting rig guide bushing contacted the refueling canal liner plate and Operating Deck curo. Radiation Protection personnel realized contact with the liner plate ano operating deck floor was going to occur, but were unable to l
communicate with the Crane Operators, and exited the airlock to the Video control Station to get help.
The Flagman entered upper containment and resumed responsibility for communications with the Crane Operators and took control of the lift.
The lower internals
DPC/MNS INPLANT REVIEW No. 91-23 PAGE 5 were placed on the stand without further incident.
4.3
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Conclusion:==
This incident has been assigned ca"ses of Inappropriate Action and Management Deficiency.
The cause of Inappropriate Action has been assigned because of less than adequate verbal communications and for failure to follow procedure.
Management ?)eficiency has been assigned as a cause becaus2 there was less than adequate management oversight, planning and work organization for the activity.
Verbal communication problems were encountered at two points during the activity. The first situation involved removal of the seal surface protective ring from the lift rig.
The potential for this problem was discussed earlier in the shift by Maintenance Engineering Services and Technical Support personnel.
It was not intended that the protective ring remain installed on the lift rig for the lower internals removal because of the affect the protective ring could have on wall clearances, core barrel target installation, and observation camera positioning.
The Maintenance Engineering Services person most knowledgeable in the work was aware of the potential impact the protective ring could have on the lift.
This was a topic of discussion with the Technical Support person prior to the Maintenance Engineering Services person leaving the station.
However, the Technical Support person
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did not have a clear understanding of the risks and consequences j
associated with conducting the lift with the protective ring installed on the lift rig and believed that he could make the lift with the protective ring installed on the lift rig.
Preparation for lifting the lower internals proceeded with the protective ring installed.
l Communications problems occurred after the initial horizontal move t
of the lower internals.
The Crane Operators and the Technical l
Support person were in contact by telephone with the Crane operators using one earpiece headsets and the Technical Support person using a telephone handset at the Video Control Station.
There were a number of people crowded in and around the Vidoo control Station who were not active participants in the lift.
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l DPC/MNS INPIANT REVIEW No. 91-23 PAGE 6 several environmental conditions existed which During this time, Radiation may have contributed to the communication problem.
monitor lEMF16 was alarming locally and the cont"! ament evacuation alarm was sounding.
(The containment evacuatiot, CArm is interlocked with IEMF16.) Neither of these audible slarms were expected to activate. The dosimetry that the Crane Operators were wearing were indicating an increasing dose rato.
(This was expected to occur as the lower internals were raised and the The Crane neutron pads moved closer to the water surface.)
operators were aware f. hat dose rates would increase based on the prejob briefing conducted with Radiation Protection persinnsi.
In addition, there was a sense of urgency to complete the lift based on discussion in prejcb meetings and the procedure directions, to The instructi.on to keep radiatisn exposure as low as possible.
begin the first horizontal movo f rom the Technical Support penson
- However, to the Crane Operator was clear and understood.
recollection of subsequent communications between the Crane Operators and the Technical Support person for subsequent. lifts indicate a after the initial internals-to-vessel contact signif icant lapse in the effective exchange of information.
Failure to follow procedure occurred several times during the Rotation of the lower internals was performed out of activity.
Installation of the elevation gauge was not done as sequence.
required by procedure. The elevation gauge was not utilized to verify that the lower internals had cleared the vessel flange.
The monitor with the underwater view of vessel flange was primarily used.
This monitor and the monitor showing the elevation gauge were physically located approximately 5 to 6 feet Radiation monitor 1 EMF 16 was not completely disabled.
apart.
- Management overaight, planning and work organization problems were identified. The management oversight function for the lift essentially ceased when the Technical Support person took over providing direction to the Crane Operators from the Video Control Station, because the Technical Support person became directly involved in performing the activity.
The Technical Support person had limited experience / knowledge performing heavy lifts and crane (He had participated in removing the lower internals flagging.
from Unit 2 in 1990, but was not the lead person for that l
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DPC/MNS INPLANT REVIEW No. 91-23 PAGE 7 activity. On Unit 2, there were 3 Maintenance Engineering Services and Technical Support persons involved in the activity when the lift was performed.)
The critical lift on Unit 1 was performed at approximately 0500 on the second night of night shifts.
Consideration had not been given to determining a contingency plan for action necessary if there was a problem once the horizontal move had been initiated.
The sequence of the lift and horizontal movements was captured on video tape.
Based upon analysis of the video tapes and the actual damage, the most probable scenario of how the damage occurred follows.
Prior to initiating the horizontal move to the deep end, the lower internals were rotated 15 degrees counterclockwise.
This rotation was performed to prevent interference of the lift rig with the Manipulator Crane. With the lower internals rotated,
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incore instrumentation guide column C-7 became the leading guide column at the elevation of contact.
on the initial impact (see pages 20 and 21), guide column C-7 moved up the side of and on top of the stud hole ledge above the 0-ring seal surface, which created the scarring on guide column C-7.
When the Crane Operator released the throttle and allowed the crane to come back to neutral, the lower internals started to twist. Guido column C-7 sheared off a small corner of the stud hole ledge.
As the internals rebounded, guide column D-10 struck the ID edge of the seal surface.
In addition, guide column E-5
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contacted the ID edge of the seal surface.
It is also believed that guide column D-10 struck the ID edge of the seal surface on a-subsequent rebound the lower internals.
On the second impact, guide column D-8 struck. Tho internals rebounded and twisted again.
During the third horizontal move, the secondary core support assembly contacted the stud hole ledge at the location that guide column C-7 contacted, then rubbed the vessel / cavity seal and bent a lifting lug approximately 90 degrees, but completely missed an excore nuclear instrumentation 1
port cover which is at approximately the same elevation as the vessel / cavity seal.
The reactor lower internals were inspected by video camera with a
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DPC/MNS INPLANT REVIEW No. 91-23 PAGE B remote operator while on the storage stand.
Several guide columne attached to the lower tie plate and the secondary core support plate showed indica', ions of contact.
The lower internals were further inspected by Framatome personnel using their photogrammetry process.
This process consists of taking several photographs at different views of the internals and using triangulation to determine angularity and/or deflections.
Both horizontal and vertical photographs of the internals were taken.
The photogrammetry data indicated guide columns C-7, C-8, D-8, D-10 end E-5 were potentially ovalized. Guide column D-10 was ovalized and bent approximately 1 to 1 1/2 degrees inward (towards the vessel centerline).
Guide column C-7 appeared dented and was severely ovalized. Guide column G-5 appeared to be slightly bent.
Damage to guide columns C-8, D-8, and E-5 involved minor scuffing and minor ovalization. The secondary core support plate was scuffed. Guide column G-5 was added to the repair list due to the Westinghouse tolerance stackup and the photogrammetry tolerances.
The lower internals were evaluated again by Babcock and Wilcox.
(B&W) using a single tube gauge, which established a vertic'al reference within the guide column and measured the col: ;.n runout.
Guide column lateral displacement and ovalarity (runout) was derived from this data.
Westinghouse Electric Company evaluated the guide column deflections to calculate the impact load applied to the reactor vessel.
This impact load was determined to be negligible with respect to other analyzed point loadings on the vessel (i.e.
seismic impact of the core barrel into the vessel-flange).
Westinghouse concurr'd with the initial video inspections that the guide columns attached to-the upper tie plate were_ unaffected since they were not impacted. Furthermore, it was concluded by analysis of the. loading and observed deflections that tue lower tie plate array was sufficiently stiff to resist plastic deformation and there was no permanent set in the array. The photogrammetry results showed all guide columns above the lower tie plate to be vertical, within the accuracy of.the analysis.
Westinghouse evaluated the effect on the impact loads on the C-7
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9 DPC/MNS INPLANT REVIEW No. 91-23 PAGE 9 quide column flange bolting and determined that two bolts (of four) may have been overstressed due to the impact. Westinghouse evaluations showed that tha remaining two bolts would provide adequate preload for normal and upset load caces, and the amount of permanent strain on the overstressed bolts was negligible.
Additionally, the impact load was used in conjunction with the observed guide tube deflections to calculate the stresses in non-impacted tubes.
This calculation demonstrated that the lower tie plate system remained elastic and the non-impacted tubes remained in their original as-built location. Visual inspections and gauging confirmed what these calculations predicted. Additional analysis of wall thickness reductions resulting from tube repaire demonstrated that 1/16th inch minimum wall thickness was adequate over an arc of 180 degrees.
Following guide tube repairs to restore original design clearances, Lhe lower internals were set in the reactor vessel.
Three remotely operated pan only video cameras were inserted through the core support and lower core plate and into the bottom mounted instrumentation area. The guide tubes in the quadrant which impacted the reactor vessel were observed to properly clear the veseal penetrations and no changes in the polar crane load cell were observed (guide tube / bottom mounted instrumentation contact would be indicated by a sharp change in :.oad).-
The reactor vessel was inspected underwater using a video camera with long handled tools to access the affected areas. The inspection showed several contact points on the reactor vessel flange at approximately 270 degrees. Additional video was cbtained from the reactor vessel ISI, which used the Seimens/KWU Central Mast Manipulator (CMM) machine.
All points of contact were on the stainless steel cladding, with no indication of cladding breaches. There was no indication of impact to the head.
and vessel alignment key at the 270 degree location.
Additionally, there was no indication of gross structural damage.
The-indications on the reactor vessel were found to be located at the stud hole ledge (above the seal surface) and the seal surface ledge (immediately below the seal surface). No contact points were observed on the 0-ring seal surface and all raised metal on l
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DPC/MNS INPIJLNT REVIEW No. 91-23 PAGE 10 the 0-ring ID edge was acceptable as is according to Westinghouse criteria; therefore, the vessel / head sealing capability is unaffected by this incident.
No credit for the stainless steel cladding is taken in the structural analysis of the reactor vessel and the cladding was not breached; therefore, the indications are acceptable as is.
After draindown of the refueling canal and prior to cleaning the 0-ring seal surface, the reactor vessel flange impact area was visually inspected by Engineering Services Representatives.
Westinghouse provided inspection and repair criteria for the reactor vessel impact locations.
The impact locations were accessible with the upper internals installed.
The five impacts of the 0-ring ID edge had raised metal pushed vertically and towards the seal surface outside diameter. The worst case was less than 1/8 inch, which satisfied the Westinghouse "acespt-as-is" criteria.
The stud hole lodge had two impacts essentially at the same location of 270*.
The first was a " cookie cutter" shearing of the edge apparently caused by guide column C-7.
Approximately 3/16 inch of the stud hole ledge, in an arc about an inch long, was sheared off with a vertical drop of 3/8 inch.
The second impact at this it sation was a rub apparently caused by the secondary core support plate. The edge of the stud hole ledge had scrape marks in the direction of travel of the lower internals. The scrape marks were approximately 1/8 inch long for approximately 3 inches-of the edge adjacent to the " cookie cutter" shear. The stainless still cladding did not appear breached in.either~ case and both satisfied the Westinghouse non-repair criteria.
Damage to the lifting rig and the refueling canal liner plate wall occurred after the internals were moved to the deep end of the canal.
The internals were lowered prior to being rotated. This action was not in accordance with the procedure. The protective ring on the lifting rig and the guide bushing of the lifting rig contacted the operating deck and then slid down the refueling canal wall.
The reactor internals lift rig (See page 22) was inspected while l
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9 DPC/MNS INPLANT REVIEW No. 91-23 PAGE 11 placed on a miselle shield, by Duke Power Design Engineering personnel, Maintenance Engineering Services personnel and a Westinghouse field service engineer.
The visual inspection showed no damage to the load carrying members of the lift rig.
The damaged support ring serves to locate the three load-carrying members as well as the alignment bushings.
The major concern following the visual inspection was the ability to engage and lock the lift rig into the internals.
Engagement with the rotolocks was verified following the inspection.
Further inspections of the lift rig were performed to assess the condition and alignment of the load carrying members. The most highly stressed we'.de in the load carrying members were dye penetrant (PT) tested, and ti.e 1
remaining welds were inspected visually. The rott.ock asser.iblies were visually inspected.
Finally, the alignment of the load-carrying members was checked by measuring center-to-center distances of the members.
The damaged guide bushing was removed to allow the lift rig to engage the remaining two reactor vessel guide pins and upper internals storage stand guide pins.
This guide bushing provides for lift rig alignment to engage the guide pins and is not necessary for operability. Repairs of the support ring and guide bushing are anticipated for the future. All of the above inspections were completed with no evidence of damage to the load carrying members or significant misalignment concerns.
Westinghouse and Duke Power evaluations determined that no repairs were required prior to returning the lifting rig to service.
Following the inspections, the lift rig was placed on the lower internals and locked-in, and load was placed on the rig verifying proper alignment and engagement.
The damage to the refueling canal liner plate was ninor and was well above the water line. Scratches were identified at the upper portion of the liner plate. The scratches extended 3 1/2 inches from the top of the liner plate and spanned an area aporoximately 4 5/8 inches wide.
The width of the largest scratch wa3 1 5/8 l
inches wide.
The depths of the scratches were also measured.
The I
maximum depth of the deepest scratch was 0.047 inches deep.
This maximum depth was located 4 7/G inches from the top of the liner plate. At all other locations, the scratch depths were 0.034
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DPC/MNS INPLANT REVIEW No. 91 PAGE 12 inches or less.
Considering the liner plate is 3/16 inches thick stainless steel and is used as a:: impervious liner with no structural loading, the damage does not impair the functional integrity of the liner plate and the liner plate is not breached.
No repair work is required.
A review of the Operating Experience Program (OEP) Database for the previous twenty-four months prior to this event was conducted.
This is the first incident involving the removal of the reactor j
vessel lower internals.
Problems with Inappropriate Actions because of failure to follow procedures are considered to be recurring.
There were no personnel injuries, radiation overexposures, or uncontrolled releases of radioacttve material as a result of this incident.
5.
CORRECTIVE ACTIONS:
Immediate:
1)
The Flagman returned to upper containment to direct the subsequent movements made by the Crane Operators.
2)
The Control Room was notified by the Operations Superintendent to monitor the incore instrumentation room sump pumpoute for any increase in frequency.
3)
A policy was implemented to require review and approval by a Lift Advisor.of any heavy or sensitive lifts in containment. Management oversight for this activity was provided by kncaledgeable Technical Support personnel.
4)
A recovery plan was developed by station management personnel which involved the following key elements:
Inspection of the lower internals.
Inspection of the refueling canal liner plate-including planning repair of known liner plate damage.
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4 DPC/MNS INPLANT REVIEW No. 91-23 PAGE 13 Evaluation of damage to the lift rig including planning repair of known damage and investigating use of other lift rige from within Duke Power and external to the company.
Evaluation of potential damage to the reactor vessel upper internals.
Examination of the reactor vessel (integrate with 10 year ISI).
5)
The Duke Power Company Significant Event Investigation Team (SEIT) was activated.
Subsequent:
1)
Procedure MP/0/A/7150/101, Reactor Vessel Lower Internals Removal and Installation, and procedure MP/0/A/7150/43, Reacter Vessel Upper Internals Removal and Installation, were evaluated'for human factors / human performance problems by the Human Performance Excellence Team (HPET) and_the SEIT and procedure changes were made.
2)
Procedure MP/0/A/7150/101, Reactor Vessel Lower Internals Removal and Installation was changed to incorporate a contingency method to address a loss of seal / canal integrity for returning the lower internals to the reactor vessel.
3)
The provisions of SOER 91-1, Infrequently Performed Tests, were implemented for returning the lower internals to the reactor vessel.
4)
Training was conducted for Polar Crane Operators and Technical Support personnel which involved:
Practice drill of Polar Crane Operators exiting the polar crane Specific repeat back communications training
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t DPC/MNS INPLA!!T REVIEW No. 91-23 PAGE 14 Dry runs of returning the lower internals to the reactor vessel.
5)
Lower Internals Repair Photogrammetry was performed to aid in determining what repairs must be done.
Guide ce?umns C-7, C-8, D-10, and E-5 were free path gauged.
Guide columns C-7, C-8, D-8, D-lO, and E-5 were expanded to achieve circular geometry.
An electrostatic diecharge machine (EDM) was used to cut the inside diameter edge of guide columns C-7, C-8, D-8, D-10, and G-5 that might interfere with the bottom mounted instrumentation and recreate a free path.
1 The secondary core support plate was deburred to remove any potential loose metal.
The ends of guide columns C-7, C-8, D-8, D-10, and E-5 were brushed ID and OD to ensure there is no loose metal remaining.
6)
SOER 85-1, Reactor Cavity Seal Failure, was reissued to evaluate seal failure during reactor vessel lower internals movements.
7)
Lifting Rig Repairs Westinghouse and Maintenance Engineering Services personnel conducted a visual inspection' to determine damage to the lifting rig.
Dye penetrant examination of the most highly stressed welds of the lifting rig was done.
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DPC/MNS INPLANT REVIEW No. 91-23 PAGE 15 j
A visual inspection of the temaining load carrying welds of the lifting rig was performed.
A visual inspection of the rotolock assemblies (the devices that are inscrted into the core barrol and rotate to engage lifting lugs on the barrel) was performed.
Alignment of the lif ting rig was checked by measuring center to center distances of load carrying members.
8)
A meeting was held with all involved personnel to discuss the event and identify actions to prevent this type of event from recurring.
9)
Work request 601665 IAE was written to provida detailed guidance for disabling IEMF16 for reinstallation of the lower internals.
Planned:
1)
The lift rig will be repaired by removing the old donut and replacing it and shimming the alignment bushings, etc. to the alignment rig.
The protective ring actuator rods will be lengthened along with the repair.
2)
System Craft Support personnel will review this event through the Work Improvement Team Process with all Reactor Head technicians, Crane Operators, Flagmen and Technical Support personnel.
3)
A caco study lessons learned package for this event which reemphasizes procedure compliance will be developed and presented to all applicable site employeec.
4)
A McGuire site group will be formed to use the excellence program to determine a solution for the problem of procedure compliance. This group will have clearly defined expectations with members to come from
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various site groups.
I 6.
SAFETY ANALYSIS:
The impact of the lower internals on the reactor cavity seal could have a range of consequences from no effect to degradation / loss of the seal with subsequent drain down to the vessel flange (See page 23).
Several scenarios were examined in response to this event.
The first scenario involved the lower internals scraping across the top of the seal pressing it tighter into the slot between the vessel flange and the floor of the liner plate (See pages 21 and 25).
This is what is believed to have happened during the incident based on the physical evidence of the position of the seal af ter the incident. Mock-up testing was also conducted to confirm this belief.
The mock-up testing consisted of a full scale reproduction of the cavity seal void eight feet in length. An eight foot section of a spare seal w2s installed and inflated. To simulate the internals being pulled across the seal, a ten ton hydraulic ram was used to force a beveled plate (simulating the leading edge of the base plate) over the seal. As suspected, this forced the seal tighter into the void.
The second scenario involved the lower internals impacting the seal at full trolley speed, dislodging the seal, pulling it partially out, pushing it across the canal floor, and then the seal returning to the slot. This scenario is credible, but the probability is very low.
The mockup was reconfigured to simulate the second scenario by forcing a vertical plate across the seal. This forced the seal up and out of the void. As expected, with the force removed, the seal' returned to esal the void.
This test was perforred with the seal both inflated and deflated with the same results.
Toe final scenario involves the lower internals contacting the nuclear instrumentation well cover at full trolley speed and completely displacing the covar. This scenario is credible, but the probability is sery low.
The dose assessment from the lower internals, if they were completely uncovered and on the storage stand, was evaluated. On the operating deck in the vicinity of the lower internals, the dose rate varies from
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DPC/MNS INPLANT REVIEW No. 91-23 PAGE 17 approximately 3300 R/hr to 2500 R/hr.
Seatter dose is estimated as 300 R/hr on the operating deck away from the canal.
Dose rates in lower containment would be less than 100 mR/hr.
There is potential for limited streaming out of the reactor building through the equipment hatch.
No dose concerns were identified for any areas of the Auxiliary Building including the Control Room, or Doghouses.
For the final scenario, with the nuclear instrumentation well cover removed, a total time of 127.3 minutes would elapse before the refueling,
canal would be drained down to the reactor cavity flange level.
Assuming that movement continued until the internals were placed on the stand (approximate duration of twelve minutes), dose rates would not begin to increase until 34.8 minutes into the drain down. At that time, the highest dose rate would be 15 R/hr on the operating deck.
Dose rates in the crane cab would be 1 R/hr.
These conditions would allow j
adequate time for the crane operators to exit containment without receiving significant dose.
Under the actual plant conditions that existed during the event, there is conclusive evidence that there was not substantial potential for overexposure. Total dose for the activity was 180 mR.
The reactor vessel, reactor internals and reactor internals lift rig were acceptable for continued operation without further repair or examination.
The impact loads caused by this incident were evaluated with respect to reactor vessel stresses and vessel fracture mechanics were shown to be within all code allowances.
Impact loads and deflections were evaluated for the lower internals bolting, stainless steel tubing and secondary core support assembly and were found to be acceptable. The lower internals have been repaired to ensure vessel penetrations will not ir.teract with the lower internals. The lower internals were reinstalled without incident.
The lift rig was rigorously inspected and no evidence of damage or misalignment to load-carrying members was found.
This incident had no impact on nuclear safety from a core safety or spent fuel. standpoint.
Spent Fuel Pool inventory and Spent Fuel cooling I
were unaftected by the incident.
Fuel Transfer Tube Valve, 1KF122, was l'
closed upon completion of defueling, maintaining separation of Spent Fue.1 Pool and refueling cavity inventories.. Refueling cavity-integrity was not lost.
This incident did not affect the health and safety of the public.
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DPC/MNS INPLANT REVIEW No. 91-33 PAGE 18
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DPC/MNS INPLANT REVIEW No. 91-23 0
PAGE 19 Event Description October 8,1991 ( Approximate 10 minute duration from ~ 0500
~ OSIO )
Lower Internals Flange breaks water surface.
Containment Evacuation Alarm sounds.
Target #4 out of water.
Target #3 out of water.
Internals rotated to clear manipulator cranc.
Target #2 out of water.
Initial Trolley horizontal travel.
Initial contact.
First rebound.
Second rebound.
Second Trolley horizontal travel.
Third rebound.
Top of larget #1 out of water.
Third Trolley horizontal travel.
Vessel Flange cleared, contact with cavity seal and lifting lug, cleared NI cover.
Trolley at storage stand - centerline mark.
Internals lowered in deep cnd.
Bridge at storage stand centerline mark.
Protective ring contacts liner plate.
Guide sleeve contacts floor and curb.
Guide sleeve slips off floor and curb.
EMF 16 and Containment Evacuation Alarms clear.
- NOON Duke Power Company Significant Event investigation Team ( SEIT ) activated.
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DPC/MNS INPLANT REVIEW No. 91-23 PAGE 20 4
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