IR 05000528/1996001
| ML17312A814 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 06/11/1996 |
| From: | Brockman K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17312A813 | List: |
| References | |
| 50-528-96-06, 50-528-96-6, 50-529-96-01, 50-529-96-1, 50-530-96-01, 50-530-96-1, NUDOCS 9606180581 | |
| Download: ML17312A814 (57) | |
Text
0 ENCLOSURE 2 U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-528/96-06 50-529/96-06 50-530/96-06 Licenses:
NPF-41 NPF-51 NPF-74 Licensee:
Arizona Public Service Company P.O.
Box 53999 Phoenix.
Arizona Facility Name:
Palo Verde Nuclear Generating Station.
Units 1. 2.
and
Inspection At:
Wintersburg
~ Arizona Inspection Conducted:
March 26 through May 22.
1996 Inspectors:
Lawrence E. Ellershaw.
Reactor Inspector.
Maintenance Branch Division of Reactor Safety John G. Kramer. Resident Inspector.
Project Branch F
Division of Reactor Projects Dale A. Powers.
Chief. Maintenance Branch Division of Reactor Safety Michael P.
Shannon.
Radiation Specialist.
Plant Support Branch Division of Reactor Safety John E. Whittemore.
Reactor Inspector, Maintenance Branch Division of Reactor Safety Approved:
nnet
roc n,
iree or Division of Reactor Safety 6 (fa(pc ITaae Ins ection Summar Areas Ins ected Units
2 and
Followup of previous inspection findings.
Areas Ins ected Unit 2 onl
Reactive.
announced inspection to followup on the stuck Fuel Assembly A07.
9606180581 960611 PDR ADOCK 05000528
i
-2-Results Unit 2 Onl P1ant 0 erations
~
Plant review board (onsite committee) meetings were conducted as thorough examinations.
with a high degree of self-critical behavior exhibited.
The boards demonstrated a thorough consideration of compensating actions to minimize the potential for fuel damage in freeing Fuel Assembly A07 from the core support plate (Section 2.3).
~
A senior reactor operator (limited to fuel handling) gave instructions that led to the application of excessive lifting force on Fuel Assembly A07.
As a result the maximum lifting load allowed by the procedure was exceeded.
This was identified as a noncited violation (Section 2.4.Z).
~
The prejob brief conducted prior to the successful attempt to free the assembly was comprehensive and stressed the proper precautions (Section 2.5.5)
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The licensee's staff employed extreme caution and conservatism in freeing and transferring Fuel Assembly AO? to the spent fuel pool (Sections 2.5.5 and 2.6).
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A senior reactor operator (limited to fuel handling) gave instructions that led to the freed fuel assembly weight on the end cap fixture to increase beyond the procedural limit.
- This was identified as a noncited violation (Section 2.6).
~
Procedural controls for handling the upper guide structure were inadequate to assure the design basis was ~aintained.
This was identified as a violation (Section 2.7.2).
~
The technical and integrated prejob briefs conducted for the lift.
transfer, and installation of the upper guide structure were noteworthy.
Extra care.
including the use of under water cameras was demonstrated in the process (Section 2.8.3).
Maintenance The electric.discharge machining of relief cuts on Fuel Assembly A07 lower fitting legs was planned and performed well (Section 2.5).
The planning. fabrication.
and installation of the alternate rigging assembly.
support strap mechanism.
wire rope mechanism.
temporary storage stand, and end cap fixture were well done (Section 2.5.3).
The electric discharge machine repairs on the upper guide structure were well planned and performed in a conservative and thorough manner with appropriate verification of machining.
where necessary (Section 2.8. 1).
-3-En ineerin The
CFR 50.59 evaluations of the activities related to the stuck fuel assembly event were comprehensive and technically substantial (Section 2.3).
Engineering designs of the alternate rigging assembly.
support strap mechanism, wire rope mechanism.
temporary support stand.
and end cap
.
fixture were innovative pioneering products.
Validation of the designs to meet performance requirements was conservative (Section 2.5.3).
The examination conducted to identify and assess the damage and determine the needed repairs to the upper guide structure was thorough and conservative (Section 2.7.2).
The likelihood of additional damage to the upper guide structure was significantly reduced because of the procedure changes made prior to the its reinstallation into the vessel (Section 2.8.2).
The onsite representative from the nuclear steam supply system manufacturer provided excellent technical assistance to the licensee in its preparations to free Fuel Assembly A07 from the core support plate (Section 2.5.5).
Plant Su ort The licensee performed accurate radiati'on dose calculations to assess the radiological implications should fuel rod failure have occurred during the attempts to free Fuel Assembly A07 (Section 2.5.4).
There were clear radiation exposure permits. with appropriate radiological controls and hold points (Section 2.5.4),
Radiation protection personnel had appropriately positioned continuous air monitors and staged air samplers on the refueling bridge and personnel evacuation routes to assess airborne dose in the event of fission gas release (Section 2.5.4).
Procedural guidance for the use of and source response check of continuous air monitors and area radiation monitors was lacking (Section 2.5.4).
The effort to free the stuck Fuel Assembly A07 was well supported by the radiation protection technical staff.
The calculations and recommendations to limit workers radiation exposure were appropriate (Section 2.5.4).
'l lt I
f l
e
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Nuclear Assurance oversight made positive self-assessment contributions by identifying the need for better communication between day and night shifts (Section 2.6).
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The root-cause team considered many potential causes and systematically eliminated numerous possibilities.
Their effort was well directed and efficient (Section 2.7. 1).
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The interim corrective actions of repairing the upper guide structure and revising the procedure for handling the upper guide structure provided assurance that the structure could be correctly installed and perform its design function (Section 2.8.3).
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The licensee-proposed long-term corrective actions appeared to be appropriate.
NRC will followup on the licensee's long-term corrective action during the next defuelings of those units (Section 2.9).
Mana ement Overview
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Inspectors observed senior management directly involved in the oversight of the operation to safely transfer Fuel Assembly A07 to the spent fuel pool (Section 2.6).
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Management and supervision were well represented in containment during the reinstallation of the upper guide structure (Section 2.8.3).
Summar of Ins ection Findin s:
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Violation 529/9606-01 was opened (Section 2.7.2).
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A noncited violation was identified (Section 2.4.2).
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A noncited violation was identified (Section 2.6).
~
Violation 528/9435-01:
529/9435-01:
530/9435-01 was reviewed.
but remains open (Section 3. 1).
~
Licensee Event Report 530/94-08 was reviewed.
but remains open (Section 3.2).
Attachment:
~
Attachment
- Persons Contacted and Exit Meeting
e
-5-DETAILS
PLANT STATUS During the inspection, Unit 1 was taken from full power to cold shutdown due to high vibration of a reactor coolant. pump.
and return subsequently to full power.
Unit 2.
was in the refueling mode until early May and then returned to full power.
Unit 3 remained at or near full power throughout the inspection.'
FOLLOWUP
-
MAINTENANCE (92902)
2. 1 Introduction
"
On March 22, 1996. with Palo Verde, Unit 2. in a refueling outage.
the licensee experienced difficulty in removing fuel assemblies located in the southeast quadrant of the core.
Adjacent Fuel Assemblies A06. A07. A08.
and 807 could not be removed with the refueling machine's cut off load limit of 1600 lbs.'ubsequently.
the licensee continued with defueling operations in the remainder of the core.
and used an underwater video camera to investigate the condition of the stuck fuel assemblies.
With the video camera.
the licensee's staff observed damage to Fuel Assembly A07.
During Cycle 6. this fuel assembly had resided in an unrodded core location in a peripheral row against the shroud wall.
The observed damage was that the top of the fuel assembly was approximately 0.7 inches lower than the adjacent assemblies.
In additions the lower end fitting on Fuel Assembly A07 was damaged, with its corners splayed outward into the fuel alignment pins.
This had caused the fuel assembly to be stuck to the core support plate.
The licensee's staff also observed approximately seven fuel rods protruding from the bottom of the fuel assembly.
through the lower end fitting.
No evidence of fuel rod cladding fai'iure was observed and this was supported by radiochemistry results of the reactor coolant.
No damage was initially visible on Fuel Assemblies A06. A08. and 807.'he licensee was eventually successful in removing these fuel assemblies once all other fuel assemblies were off loaded to the spent fuel pool.
The spacer grids on Fuel Assembly A07 were observed to be lower than adjacent assembly spacer grids.
The licensee formed task groups to develop plans for removing the last fuel assembly.
to assess radiological consequences of fuel rod failure. to recage the fuel assembly.
to establish alternate means for lifting the fuel assembly.
to inspect and assess damage to the upper guide structure.
and to determine a
root cause for the damage.
As a convenience.
in this inspection report fue> assemblies ~ill be identified by the core lattice locations that they occupied during Cycle 5 operatio it e
e-6-The inspectors were not aware of any other domestic events similar to the one at Palo Verde where a fuel assembly became stuck to the core support plate.
2.2 Fuel Assembl U
er Guide Structure and La down Pad Desi ns The fuel for Palo Verde Nuclear Generating Station was of the Combustion Engineering Standard Safety Analysis Report design.
Within a fuel assembly.
the fuel rods were arranged in a
X 16 array.
Each fuel assembly contained 236 fuel rods and 5 guide tubes.
The array was held together by 10 Zircaloy-4 spacer grids welded to the guide tubes and a bottom Inconel spacer grid welded to the lower end fitting.
At the top end of the fuel assembly.
the guide tubes were connected to guide posts.
which attached to the flow plate.
holddown springs.
and the holddown plate.
At the lower end of the fuel assembly.
the guide tubes were connected to the lower end fitting.
The lower end fitting had four support legs.
Fuel rods.
guide tubes.
and guide tube posts were constructed of 2ircaloy-4.
The fuel assembly's center guide tube served as a conduit for an incore instrument that entered from the bottom of the fuel assembly through a funnel-shaped nozzle.
The proper alignment of fuel assemblies on the core support plate was achieved by positioning the corners 'of each fuel assembly against four alignment pins.
Except on the core periphery.
each alignment pin contacted four fuel assemblies.
Alignment was provided by the indented shape, of the corners of the lower end fitting support legs.
which matched up to the alignment pins.
The upper guide structure was comprised of a fuel alignment plate and a
support barrel assembly through which control element assemblies were housed and properly aligned.
Extending down below the fuel alignment plate were guide tubes'hich provided a conduit for inserting the control element assemblies into the fuel assemblies.
Mhen properly placed, the upper guide structure's guide tubes encompassed the uppermost part of the fuel assembly guide posts.
The laydown area for the upper guide structure was at the east end of the reactor cavity.
In this area.
four separate laydown pad assemblies located
degrees apart.
provided upper guide structure storage by directly supporting the fuel al.ignment plate during refueling operations.
Four slots in the fuel alignment plate were designed to engage tapered risers that extended about 9-1/2 inches above the laydown pad assemblies.
Continued lowering of the upper guide structure onto the tapered risers would correctly orient the structure on the laydown pads.
The inspectors noted that the guide tubes extended about 10-1/2 inches below the fuel alignment plate.
2.3
CFR 50.59 Evaluations and Plant Review Boards The inspectors reviewed several
CFR 50.59 screenings and evaluations related to the circumstances associated with freeing the stuck fuel assembly.
They also observed several plant review board meetings convened by the licensee to review the
CFR 50.59 screenings and evaluations of the various issues.
The inspectors observed that these meetings were all conducted
i i
(
k l
I
formally and a quorum was present.
The inspectors observed the conduct of these meetings was self critical and conservative in the critique of issues.
Additionally. the plant review boards thoroughly discussed any procedure or guidance changes from plant review board issues previously approved.
The discussion below relates to a typical evaluation that was subsequently accepted by a p1ant review board.
The boards demonstrated a thorough consideration of compensating actions to minimize the potential for fue1 damage in freeing Fuel Assembly A07 from the core support plate.
The inspectors reviewed the
CFR 50.59 evaluation for changing the radiation monitor HI alarm setpoints for RU-16 (140 ft containment operating level) and RU-33 (refueling machine area)
during Fuel Assembly A07 removal.
An inspector discussed the changes with licensee personnel.
Licensee personnel noted that the basis for the HI alarm setpoint was to inform workers when radiation levels on the refueling floor reached 2.5 mrem/hour (this was in accordance with the updated final safety analysis report).
The KI alarms were the threshold values for initiating a declaration of an ALERT emergency classification.
The licensee determined that with the normal alarm setpoint values.
the classification would be extremely conservative during the Fuel Assembly A07 retrieval.
The licensee raised the HI alarm setpoints to I rem/hour
. which correlated to other emergency action levels for radiological events at the site.
The temporary setpoint change correlated to the failure of approximately eight fuel rods for the specified time after reactor shutdown.
The licensee performed an appropriate evaluation for the setpoint change.
The licensee personnel exhibited a proactive approach in evaluating the effects of the fuel assembly recovery.
Based on the review of 10 CFR 50.59 evaluations, the inspectors determined that the evaluations to free the stuck fuel assembly were comprehensive and technically substantial.
The inspectors further concluded that the 10 CFR 50.59 reviews by the plant review board appropriately addressed questions and the plant review board meetings were thorough deliberations.
2.4 Second Attem t to Free Fuel Assemblies From the Vessel Once the unaffected fuel assemblies had been off-loaded to the spent fuel pool. there were four fuel assemblies left in the core.
The licensee's intent was to free the three fuel assemblies that had no visible damage.
2.4. 1 Removal of Fuel Assemblies A06. A08. and BO?
On March 24-27.
1996.
licensee personnel.
using the refueling machine and without exceeding the Technical Specification limit of 1600 lbs. were successful in removing Fuel Assemblies A06. A08. and 807 without inciden e-8-No damage was initially visible on Fuel Assemblies A06, A08.
and B07: however.
after the removal of these assemblies from their core locations, further examination was performed.
Examination of Fuel Assemblies A06 and AOB revealed significant grid strap damage on the sides adjacent to Fuel Assembly A07.
A small piece of debris.
appea.ing to be a piece of grid strap from Fuel Assembly A07 was observed on the side of Fuel Assembly B07 that was adjacent to Fuel Assembly A07.
Ultimately. the reactor engineering and fuel handling organizations performed additional inspection on all fuel assemblies that could have been impacted by the defective or deformed upper guide structure.
Particular attention was paid to Fuel Assemblies A06. A08. B07.
and TlO.
The licensee found no significant indication other than the grid strap damage on the sides of Fuel Assemblies A06 and AQ8. which were next to Fuel Assembly A07.
There was no indication that other fuel assemblies had been affected by the defective upper guide structure.
Subsequently.
Fuel Assembly B07 was reinstalled into the core for the next operating cycle.
2.4.2 Second Attempt to Free Fuel Assembly A07 For the second attempt to free stuck Fuel Assembly A07
~ the licensee used hydraulic jacks that were placed under the fuel assembly.
Coincident with the use of the jacks'he licensee applied lifting force with the refueling machine hoist.
Palo Verde Unit, 2 Technical Specification 3/4.9.6 stated that the refueling machine shall be used for movement of fuel assemblies and shall be operable with a minimum capacity of 359Q lb and an overload cut off limit of less than or equal to 1600 lbs.
The inspectors reviewed Procedure 72IC-9RX03,
"Core Reloading." Revision 1.
and noted that the procedure referred to the licensee contractor Procedure 13-N0001-5.05-435-1.
"Fue'. Assembly Hydraulic Lifting Tool."
Revision 1.
Prior to commencing the lift. the inspectors noted that a copy of the contractor's procedure was not available in the control room.
An inspector questioned control room personnel about the unavailability of the contractor procedure.
A reactor engineer subsequently retrieved a copy of the procedure for use in the control room.
The inspectors concluded that the control room staff demonstrated a lack of a questioning attitude by not having a copy of all applicable procedures available for reference in the control room.
An inspector in the control room observed the second attempt by the licensee to free Fuel Assembly A07 on March 28.
1996.
During preparation for the second attempt.
the inspector observed the shift supervisor communicate specific elements of the procedure to the reactor engineering communicator.
The reactor engineering communicator correctly repeated the communications back to the shift supervisor.
The communications were then relayed to the senior reactor operator (limited to fuel handling) inside containmen e During the second attempt.
an inspector observed control room person'nel receive information that refueling personnel had placed 1600 lb of hoist lift force on the assembly.
This exceeded the procedural limit specified as 50 percent of the assembly weight (about 800 lbs).
Contractor Procedure 13-N0001-5.05-435-1.
Steps 4.5 and 4.7.
were being performed pursuant to Step 20. 1 of licensee Procedure 72IC-9RX03 when the excess force was applied.
An operations management representative was present on the refueling bridge during this activity.
The shift supervisor immediately directed refueling personnel in containment to reduce the lifting force to approximately 50 percent of the assembly weight as directed by the procedure.
The hoist lift force was immediately reduced to 800 lbs.
The shift supervisor halted further attempts to free the fuel assembly in order to evaluate the situation.
The licensee initiated Condition Report/Disposition Request 2-6-0057 to evaluate and provide corrective action to preclude recurrence of the violation.
Licensee personnel asked the contractor to determine and provide the basis for limiting the refueling machine hoist lifting force to 50 percent of the assembly weight.
The contractor responded that the requirement was meant to ensure that the hydraulic jacks would not slip out from under the fuel assembly.
The contractor also indicated that placing a
1600 lb lift on the fuel assembly from the refueling machine would be acceptable.
The inspectors determined thai the failure of the senior reactor operator (limited to fuel handling) to follow the procedure constituted a violation of Technical Specification 6.8. 1.
However. this failure to follow procedure was determined to be a violation of minor significance and would be treated as a
noncited violation. consistent with Section IV of the Enforcement Policy.
To ensure that expectations were understood and subsequently met
~ the shift supervisor held a face-to-face discussion with a new senior reactor operator on the procedure requirements for the assembly removal.
Licensee personnel continued with the attempt to free the fuel assembly using the allowed higher hoist loading limit.
The senior reactor operator reported that a banging noise was heard in the vicinity of the fuel assembly.
The refueling operators noted that the lifting force had decreased.
and the upper portion of the fuel assembly had moved up.
The licensee conservatively stopped the recovery attempt at this time to evaluate the unanticipated results.
2.5 Third Attem t to Free Fuel Assembl A07 Following the second attempt to remove Fuel Assembly A07 its structural integrity was questionable.
Therefore. it was necessary to design and fabricate a means for supporting the fuel assembly.
Since the support mechanism was to be on the outside of the fuel assembly.
there was insufficient clearance to withdraw the assembly into the refueling machine mast.
Additionally. Technical Specification 3/4.9.6 specified that only the
l I
"f
-10-refueling machine could be used for movement of fuel assemblies.
Therefore.
the licensee proposed to revise Technical Specification 3/4.9.6.
As a result.
Amendment 96 was issued.
allowing the use of alternate equipment (a manually controlled hoist) to withdraw the fuel assembly from the reactor.
Also. the licensee was aware that the fuel assembly lower end fitting legs were splayed against and wedged into fuel alignment pins on the core support plate.
The licensee.
in conjunction with ABB-Combustion Engineering.
Inc..
developed a plan to cut two relief slots into the support leg webs of the lower end fitting of Fuel Assembly A07.
This slotting would allow the support legs to deform slightly inward.
upon lifting. without compromising the overall integrity of the lower end fitting. and free them from the alignment pins.
ABB-Combustion Engineering performed finite element analysis calculations for the lower end fitting with and without the designed slots.
The analysis concluded the structural integrity of the lower end fitting would not be significantly impacted by making the slots.
The licensee's review and concurrence with the conclusions of the analysis and the process methodology were documented in Appendix T to Revision 1 of Procedure 721C-9R)(03.
The method used to cut the slots was electric discharge machining'hich is a
'rocess that vaporizes metal using electrical arcing.
ABB-Combustion Engineering implemented the process with licensee support.
ABB-Combustion Engineering used Procedure V2-NOME-EP-0100.
"EDM Contingency Step."
Revision 3.
The inspectors observed the setup of the electric discharge machining equipment.
and the first of two cuts made on the lower end fitting support legs.
The process was performed underwater with precisely controlled amperage and voltage.
The burn rate was slow (approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per cut)
~
and the resulting residue was. essentially, a fine powder.
The process set-up and cutting operations were monitored by video cameras which provided excellent resolution.
Personnel took preca'utions to protect the fuel rods that had dropped below the lower end fitting by >>acing a 1/8 inch plate between the lower end fitting support legs and the fuel rods.
The inspectors concluded that the cutting process was well planned. thorough'nd performed in a conservative manner.
2.5. 1 Alternate Rigging Assembly The alternate rigging assembly consisted of an 8.3 ft long X 4 inch square steel tube that formed a rigging beam attached to the top (front and rear) of the refueling machine hoist frame with U-bolts.
The beam extended beyond the front of the refueling machine hoist frame for 41 inches.
A lug was welded to the underside of the beam so that the load cell and hoist assembly could be attached at a position 5 inches from the end.
A calibrated load cell. which had a 5000 lb capacity.
was attached.
Engineering personnel performed Calculation 13-CA-ZC-961 to validate the design structural adequacy and loading capacity.
The inspectors reviewed the calculation and determined it to be conservative.
Engineering personnel used the maximum fuel assembly load limit of 1600 lbs.
and added a conservative miscellaneous rigging hardware weight of 400 lbs. for a total of 2000 lbs.
A design load factor of 1.5 was
t
-11-used:
thus.
the design loading capacity of the alternate rigging assembly was established as 3000 lbs.
Licensee personnel recognized that the use of an alternate rigging assembly would result in the loss of built-in electrical and mechanical protective functions provided by the refueling machine.
Consequently.
they performed a
review to identify all refueling machine interlocks and their functions.
For each of these interlocks.
mechanical and/or procedural controls were established to provide compensatory actions to prevent damage or dropping of the fuel assembly.
The alternate rigging assembly was load tested by applying a lift strain unti 1 the 3000 lb load test force was reached.
The inspectors observed the trial run of the alternate rigging assembly, in which a dummy fuel assembly was used to verify its intended performance capability.
2.5.2 External Fuel Assembly Support Since the condition of the load-bearing guide tubes of Fuel Assembly A07 was unknown. engineering personnel conservatively concluded that an exterior caging arrangement needed to be designed and installed,to provide the complete support for the fuel assembly during removal. transport.
and storage.
The exterior caging was designed and designated the support strap mechanism.
It included two straps with a J-hook arrangement.
a jacking plate device to tension the straps.
and a bearing plate.
These were fabricated onsite.
using stainless steel material.
The strap mechanisms were made from I inch wide by 1/4 inch thick stainless steel plate. while the bearing plate and jacking plate overall dimensions were approximately 8 inch square by I inch thick.
The two support straps were hooked on opposite sides to the fuel assembly lower end fitting. and the jacking plate rested on top of the bearing plate and fit on the fuel assembly upper end fitting.
The jacking plate had adjusting bolts that would press against the bearing plate and thus pull up on.
and tension.
the straps.
A concern was raised regarding possible buckling (upon tensioning) of the fuel assembly lower end fitting. If this occurred't could push the straps radially outward such that the modified fuel assembly would not fit in the upender.
Through the use of specially designed gaging. it was determined that the fuel assembly would fit in the fuel assembly upender.
but the tolerances were so close that any bowing. twisting. or misaligning of the assembly in the upender might require a greater force than the damaged fuel assembly could withstand in order to extricate it.
Therefore.
engineering personnel designed an alternate device. called the "wire rope mechanism."
The wire rope mechanism would be used to replace the support strap mechanism (once the fuel assembly was ready to be placed in the upender)
and maintain structural integrity of the fuel assembly.
Since the dimensions of the wire rope mechanism were less than the support strap mechanism.
the potential for dimensional interference in the upender was reduce l
e-12-The wire rope mechanism consisted of aircraft grade wire rope with loops at each end that were compatible with the support strap mechanism jacking plate on top of the upper end fitting.
Two of these devices were required.
The loop at one end attached to one corner of the jacking plate. while the other end was threaded down the exterior of the fuel assembly.
through the lower end fitting. back up the opposite side of the fuel assembly.
and attached to the adjacent corner of the jacking plate.
The other wire rope mechanism was similarly attached to the other corners of the jacking plate.
This in effect.
formed a cage which provided structural integrity to the damaged fuel assembly.
Appendix S of Procedure 721C-9RX03 provided the instructions for attachment of the support strap mechanism and the wire rope mechanism.
The inspectors reviewed Calculation 13-CA-ZC-961 which provided the design and loading analyses.
and observed the proof-load tests conducted on each support strap mechanism (2400 lbs)
and each wire rope mechanism (in excess of 2800 lbs) to verify integrity of design, materials'nd fabrication.
For added assurance.
licensee personnel made a decision to attach both of the mechanisms (support strap and wire rope) before the next attempted lift from the reactor core support 'plate.
This decision to remove the support strap mechanism prior to placing it in the upender created a need.
however. to construct a temporary support stand.
The support stand is discussed below.
2.5.3 Additional Equipment Developed for Safe Fuel Assembly Transit A temporary support stand was designed and fabricated to provide a rigid and stable support.
and to prevent the fuel assembly from being free standing during removal of the support strap mechanism.
The stand was fabricated from sections of 14 inch pipe and anchored in the reactor cavity on the route that would be taken to place the freed assembly in the upender.
The analysis of the temporary support stand design showed it to be functionally equivalent to other normal fuel storage locations within the containment and fuel building structures.
To limit any further movement of fuel rods that protruded into the.lower end fitting during removal and transport of the fuel assembly.
a stainless steel end cap fixture was developed to act as a new lower end fitting for the bottom of the fuel assembly.
The center of the fixture was designed with a counterbore to allow it to support the weight of the fuel assembly, without bearing on the fuel rods that had dropped down below the fuel assembly support legs.
Flow holes were provided in the end cap fixture to allow the cooling flow normally present in the fuel assembly when placed in its storage location.
The fixture was placed on a larger support plate and lowered down onto the top of the core support alignment pins next to the stuck Fuel Assembly A07.
Once lifted and in position, Fuel Assembly A07 was to be lowered onto the fixture.
The design of the fixture was such that it positively attached to the lower end fitting of the fuel assembly through the use of four spring-loaded.
swivel-locking arms that were activated by the weight of the fuel assembl t I
-13-The licensee performed a conservative analysis of the end cap fixture design using a dynamic factor of two times the mass of the fuel rods in the assemb1y to ensure that it would not fail during transport of the fuel assembly or during storage in the spent fuel pool.
The placement of the fuel assembly onto the end can fixture was ana1yzed in Appendix P of Procedure 72IC-9RX03.
The inspectors concluded that the design, fabrications planning.
and installation of the a1ternate rigging assembly.
support strap mechanism.
wire rope mechanism.
temporary storage stand.
and end cap fixture were well done.
and they resulted in innovative pioneering products.
Validation of the design to meet performance requirements was conservative.
2.5.4 Radiological Considerations The inspectors reviewed calculations performed by the licensee to project the airborne dose rates in containment if fuel rod failure were to occur when the stuck fuel assembly was lifted from the core support plate.
Additionally. the inspectors independently performed such calculations.
Following the onsite portion of the inspection.
the inspectors reviewed additional dose rate calculations dated May 2.
1996.
and based on 6.5 days after shutdown.
This time coincided with the attempted initial removal of Fuel Assembly A07.
The inspectors determined that the licensee's calculations were appropriate.
The inspectors.
furthermore.
determined that if a fuel rod was to break.
the operating crew would be able to place the damaged fuel assembly in a safe condition and leave containment without exceeding regulatory dose limits.
The inspectors determined that radiation protection personnel had identified the proper predominate radiological isotopes that would be released to the atmosphere in the event of fuel rod failure and the associated hazard with each isotope.
The licensee calculations were conservative and recommendations to limit exposure were appropriate.
Recommendations to limit exposure included equiping personnel who would remain in containment during.the lifting and transporting of Fuel Assembly A07 with plastic clothing and respirators.
The inspectors reviewed the following radiation exposure permits which were used for the preparation and retrieval of the stuck spent fuel assembly:
~
REP 2-96-3021B.
"Retrieve Spent Fuel Assembly A07":
REP 2-96-3024B.
"Engineering Evaluations.
Preparation.
and Underwater Video Inspections in Support of UGS Repairs and Retrieval of Fuel Assembly A07"; and.
~
REP 2-96-3026A.
"Perform EOM on Lower End Fixture Leg of Fuel Assembly A07."
The inspectors determined that the above referenced radiation exposure permits were clear and concise.
Radiological controls.
hold points. job coverage and
e-14-prejob briefing requirements were defined and appropriate for the tasks covered by the permits.
The inspectors performed walk-downs of the containment and fuel buildings to verify proper positioning of radiological monitoring equipment in support of the fuel assembly retrieval effort.
The inspectors noted that the licensee had assigned monitoring locations on the basis of the airflow direction in the containment and fuel building during the planned retrieval and transport.
In addition to appropriately positioning continuous air monitors'he licensee staged job coverage air samplers on the refueling bridge and personnel evacuation routes to assess airborne conditions in the event of a fission gas release.
The inspectors verified the operation of portable continuous air monitors and area radiation monitors located in the containment and fuel buildings.
The inspectors reviewed the instrument response check record stickers attached to the continuous air monitors and determined they had been appropriately source checked.
However. it was also noted that the continuous air monitor strip chart recorder clocks for the containment and fuel building varied as much as 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> from actual time.
The inspectors reviewed Procedure 75RP-9EE07.
"AMS-3 Monitor Operations."
Revision 5.
and noted that radiation protection personnel assigned to an area.
were not required to verify that the strip charts were indicating the actual clock time.
In discussions held with licensee radiation protection management.
an inspector pointed out the advantages of event reconstruction and assessment if actual time was indicated by a monitor.
The inspectors noted that the instrument response check record for a containment continuous monitor was initialed for April 2, 1996.
A licensee representative was questioned as to why there were initials but no results on the instrument response check record.
The representative stated that signing initials indicated satisfactory performance and that was the same as recording the resu1ts.
When the inspector discussed the above issue with radiation protection management the representative stated that their practice had been an oversight and actions would be taken to correct the inconsistency.
The licensee's staff appropriately initiated Condition Report/Disposition Request 96-0-347 to address the monitor clock settings and the inconsistency issues.
An inspector requested the licensee to source response check the continuous air monitor in the fuel building.
A senior radiation protection technician performed this source response check.
The containment air monitor failed low when source response checked.
The radiation protection technician took the proper actions to place the monitor out of service.
The inspector reviewed the air monitor strip chart for April 2.
1996.
and could not identify an indication on the strip chart that the continuous air monitor had been source response checked.
or that it had passed the source response check.
Licensee representatives questioned by the inspector stated that the strip chart appeared to be functioning properly.
Radiation protection management was not
-15-able to explain why the source response check did not show up on the strip chart.
when the record indicated that it had been performed.
During a followup conversation with licensee radiation protection management personnel.
the inspector was informed that the technician who source response checked the AHS-3 monitor in the fuel building believed he locked the marking bar during the source response check.
which prevented the chart from being marked.*
The inspector noted that there was no procedural guidance for personnel to lockout this bar when source response checking this equipment.
The inspector concluded that locking out the striker bar when source response checking this equipment was not needed to perform a source response check.
Additionally, without procedural guidance concerning the proper method to lock out the striker bar. it would be possible to improperly restore the equipment.
The inspector verified the operation of the area radiation monitors located in the containment and fuel buildings.
The senior radiation protection technician assigned by the licensee was not familiar with the operation of the area radiation monitors and, therefore, was not able to demonstrate the setting of the alert and alarm levels on this equipment.
The inspector determined that there were no operational procedures or guidance available to aid personnel. in the operation of these monitors.
Licensee personnel stated that.they would call the instrument calibration laboratory to get verbal direction when setting this monitor.
The licensee's staff appropriately wrote Condition Report/Disposition Request 96-0-347 to address this issue as well as the issues discussed above.
In summary.
the inspectors concluded that air sampling equipment was appropriately positioned to assess airborne conditions in the event of a radiological incident.
However. guidance for the use and source response check of continuous air monitors and area radiation monitors needed improvement.
It was also concluded that the licensee effort to free the stuck assembly was properly supported by the radiation protection technical staff.
The licensee calculations and recommendations to limit workers'adiation exposure were appropriately and task-related radiation exposure permits were easy to read with appropriate radiological controls.
2.5.5 Freeing Stuck Fuel Assembly A07 Prior to attempting to free the stuck assembly.
fuel handling personnel installed the external fuel assembly support mechanisms described above.
The lower end cap fixture. however.
would be installed on the bottom of the fuel assembly after it was free.
On April 7.
1996, the inspectors attended the prejob briefing for the removal operations.
All personnel involved in the task were identified and job assignments were discussed prior to discussing the actual details of the task.
The inspectors noted that the work group leaders discussed the tasks in detail.
asked all workers involved for questions.
and stressed industrial safety as a requirement for successful results.
Operations management stated that all personnel involved in the tasks had stop work authority.
The
I t
(
e-16-radiation exposure permit, as low as reasonably achievable requirements.
hold points.
and radiological survey data were discussed with involved personnel by the radiation protection section leader.
Additionally. the radiation protection section leader stressed the radiological protection actions to take if a continuous air monitor or area radiation monitor alarmed and the potential radiological exposure rates if fuel rod failure were to occur.
The inspectors observed that the onsite representative from the nuclear steam supply system manufacturer was helpful and provided excellent technical assistance to the licensee in its preparations to free Fuel Assembly A07.
His involvement included advising on the potential behavior of the fuel assembly upon its liftofffrom the core support plate.
The inspectors concluded that the prejob briefing was comprehensive and stressed proper precautions.
Later on April 7, 1996.
the licensee evacuated containment of nonessential personnel.
An inspector remained in containment on the refueling bridge to witness the retrieval of the stuck Fuel Assembly A07.
The inspector used Plant Procedure 72IC-9RX03.
"Core Reloading." Revision 5. to follow the process.
The inspector observed that all personnel knew their assigned tasks.
were alert to potential hazards.
and performed in a professional manner during the entire evolution.
In general.
the senior reactor operator maintained proper control of the operation.
The inspector noted that when the hydraulic control line of a jack became twisted in the chain hoist line attached to the fuel assembly.
the senior reactor operator solicited and evaluated ideas from personnel on the crew and in the control room before proceeding to untangle the lines.
Radiation protection personnel continuously monitored the radiological conditions, to ensure workers were not unnecessarily exposed.
The radiation protection section leader stopped work when he recognized a
potential industrial safety hazard.
The inspector noted the potential hazard was professionally addressed and resolved'nd activities, resumed.
The refueling personnel and contract personnel began applying jacking and hoisting in accordance with the procedure.
Before any procedural limits were approached, Fuel Assembly A07 came free from its stuck position on the core support plate at 5: 15 a.m.. April 7.
1996.
Immediate inspection revealed that there was no significant change in assembly integrity or configuration.
Monitoring equipment and radiological surveys confirmed that there had been no breach of fuel rod cladding.
The inspectors concluded that the licensee had conducted the operation to free the fuel assembly in a conservative.
safe manne.6 Movement of Fuel Assembl A07 to Safe Stora e
Once Fuel Assembly A07 was freed from the core support plate.
the licensee began installation the end cap fixture. the device described above to assure that fuel rods protruding through the broken lower end fitting did not fall out of the fuel assembly.
The fixture. also referred to as the "bear claw."
was designed with swing hooks that would automatically attach to the assembly.
when the assembly weight was placed on the device.
The end cap fixture. wire rope mechanism.
and support straps would assure that the assembly retained its configuration as it was transferred to a temporary storage stand.
before placement in the fuel assembly upender.
This operation to install the end cap proved to be difficult and time-consuming.
The inspectors observed that successful placement of the fuel assembly into the end cap fixture took over 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
Inspectors noted that Plant Procedure 72IC-9RX03.
Step 45.0 stated.
"Lower the fuel assembly onto the special end cap fixture.
Maintain at least 1000 lb indicated load on the dynamometer."
Ouring an effort to latch the end cap fixture on the fuel assembly, an inspector observed the senior reactor operator (limited to fuel handling) directed the fuel assembly to be lowered onto the special end cap fixture to an indicated load on the dynamometer of 850 lbs.
The inspector questioned the reported indicated load reading with the operations management oversight supervisor on the refueling bridge.
The senior reactor operator was cautioned by the operations management oversight supervisor about the 1000 lb'inimum step in the procedure and the proper tension was then placed back on the fuel assembly.
The actual safety consequences of lowering the fuel assembly, in this case to an indicated load on the dynamometer of 850 lbs.
was minimal.
Inspectors later determined that limit had been initiated and approved by the plant review board.
during review of the procedure.
The board did not identify a basis for the minimum hoist load.
However.
the issue is of regulatory concern since the senior reactor operator did not realize he had exceeded procedural guidance.
until corrected by the operations management oversight supervisor.
The inspectors concluded that, this failure to follow procedure constituted a violation of minor significance and would be treated as a
Non-Cited Violation. consistent with Section IV of the Enforcement Policy.
In resuming work. fuel handling personnel exercised patience and were successful in attaching the end cap fixture.
Personnel used extreme caution and conservatism in transferring the freed assembly to the intermediate storage stand.
The inspectors observed senior management directly involved in the oversight of this phase of the fuel assembly transfer.
Once the fuel assembly was in the temporary storage stand.
the support straps were removed to facilitate further transfer and storage.
The ai rcraft cables were verified at the proper tension and the containment fueling crew was changed.
The new crew carefully extracted the fuel assembly from the temporary storage stand and placed it in the fuel assembly upender.
The assembly was transferred to the fuel building via the transfer cart and safely
e-18-placed in a quarantined storage area of the Unit 2 spent fuel pool at about 3:40 a.m., April 8.
1996.
Nuclear Assurance provided round-the-clock coverage of activities related to the removal of the stuck assembly.
The inspectors reviewed Nuclear Assurance Evaluation Report ER-96-0225.
which provided independent evaluation of in-process activities related to removal of the damaged fuel assembly.
The report identified a weakness in the shift turnover between day and night shifts. which contributed to delays in the final design of the fuel assembly removal equipment.
The inspectors believed that the findings and recommendations that resulted from the evaluation were positive and beneficial to the licensee's total effort.
The inspectors concluded that generally.
the licensee handling and transport of the fuel assembly was cautious and performed with a high regard for plant personnel and public safety.
Senior management provided good oversight during the transfer.
2.7 Dama e Assessment and Root Cause Determination The licensee formed separate teams to identify and assess any damage to reactor vessel internals and to identify the significant contributors and the root cause of the fuel assembly becoming stuck.
These teams worked separately to achieve the stated goals.
Once the damage was identified and assessed'he assessment team's focus shifted to repairing the damaged upper guide structure.
2.7. 1 Reactor Vessel Internals Damage Assessment This team.
formed to identify. assess.
and repair any damage to the reactor vessel internals.
consisted of engineering personnel under a mechanical design engineering supervisor.
with responsibility for contractor interface and oversight of contractor activities.
The initial goal of the team was to identify any damage to the lower core support structure.
fuel assembly alignment pins.
and upp r guide structure.
As previously described a complete visual inspection of the relevent components was performed with underwater video cameras.
The inspection of the upper guide structure revealed several indications of possible damage to the bottom section of the guide tubes, which normally matched up with and fit over the fuel assembly guide posts.
As a result of these indications'he licensee performed additional inspections.
The only indications on the lower core plate were markings left by the jacks used to free the stuck assembly, The visual inspection of the fuel assembly alignment pins did not reveal any damage indications.
Additionally. a gauge block was used on the alignment pins to check pin dimension.
identify any pin deformation.
and assure correct pin orientation.
There was no detected damage to the fuel assembly alignment pin The licensee's contractor inspected and identified the damage to the upper guide structure lower guide tubes using robotic equipment and special tooling.
Identification and assessment of the damage was performed using contractor procedures.
The inspectors verified that the procedures were subjected to the licensee's normal process for review and approval of contractor procedures.
For the additional inspections of the guide tubes on the upper guide structure, a single-pin gauge block was as used to establish a minimum diameter in the lower part of each tube.
Using this gauge. it was determined that all guide tubes met the minimum diameter requirements except for the one tube that was positioned over Fuel Assembly A07.
Subsequently.
additional gauging tools were developed and gauging was performed using four-pin and six-pin gauge blocks.
Successful gauging with the four-pin block assured that the proper configuration existed for the four tubes within a set that corresponded to a single fuel assembly.
The six-pin gauge block provided assurance that there was proper orientation between sets of tubes for adjoining fuel assemblies.
The multi-pin gauging effort indicated that there were configuration and orientation problems with one guide tube over Fuel Assembly T10 and two guide tubes over Fuel Assembly A08.
The final assessment of the upper guide structure indicated that four guide tubes needed to be repaired.
The inspectors concluded that the licensee was thorough and conservative in the approach to identifying and assessing the damage to the upper guide structure.
2.7.2 Root-Cause Determination The inspectors concluded that the root-cause investigative team was established and worked according to licensee procedures specifically intended for investigating events.
determining the root cause.
and implementing the proper corrective action.
The inspectors reviewed the procedures used for event investigation and cause determination.
Procedure 90DP-OIP07.
"Significant Condition Investigations
" Revision 2.
was the basic procedure used to determine the root cause of the stuck fuel assembly.
The inspectors verified that the designated team leader met the training and qualification requirements specified by the procedure.
The inspectors were unable to verify that the indivi'dual team members were qualified as specifi,ed in the procedure.
Procedure Step 3. l. 16 stated that individual team members should be qualified in accordance with ANSI/ANS 3. 1 (1978) for their discipline.
The inspectors reviewed ANSI/ANS 3.1 (1978)
Paragraph 4.4.1-4.4.5 and determined that the required qualifications consisted of academic and experience requirements for the technical professional disciplines in reactor engineering.
radiochemistry, radiation protection.
instrument and controls.
and quality assurance.
The inspectors asked a licensee representative to explain how this requirement was met and provide documentation of this qualification for each team member.
The representative told the inspector that there was no documentation of such qualification and that the requirement appeared meaningless in the context of
-20-the procedure.
He indicated that the requirement had been placed in the procedure erroneously and would be removed.
The inspectors observed that removing this requirement would effectively eliminate any qualification for an individual to be a member of an investigative team.
However. there were adequate requirements to assure qualification of the team leader.
Regardless of this uncertainty.
the inspectors identified no team members who appeared unqualified for their assignment.
Procedure 90DP-OIP07 was supported by detailed guidance found in Procedure 90IG-OIG01.
"Root Cause Investigation Manual." Revision 0.
The Root Cause Investigation Manual was newly developed and approved guidance.
and was under the control of Nuclear Assurance and not controlled by the site document control system.
The inspectors reviewed the manual and noted that it contained the industry-accepted investigative process and guidance for various analytical methods and techniques.
Two discussions were held with the team regarding their progress in identifying the root cause.
The inspectors had numerous other contacts with the team leader to assess progress.
The inspectors noted that the team considered many avenues and systematically eliminated several possibilities for the root cause.
The team used the event and causal factor charting approach.
as delineated in the Root Cause Investigation Manual.
During discussions and interactions with the team, the inspectors observed the root-cause team to be effective in communicating and interacting with other site organizations.
The team's activities were well directed and efficient.
The inspectors also observed the team to be effective in identifying the existence of and retrieving informat~on needed to identify the cause of the event.
Prior to reaching a final decision on the root cause.
the team made some important findings and recommendations.
IJ The cause of Fuel Assembly A07 damage appeared to be attributable to damaged upper guide structure guide tubes.
This damage had occurred because of impacts against the lay down pads, during placement or removal.
As a result of this damage.
Fuel Assembly A07 was. in turn.
damaged during installation of the'pper guide structure in the reactor vessel.
A preliminary determination was reached that the problem was applicable to Units 1 and 3.
The current design of hardware for correct alignment of upper guide structure over laydown pads was probably inadequate to preclude future damage.
and a design change would be needed.
A procedure change was needed to assure that the upper guide structure could be transported from the laydown pad assemblies to the vessel without incurring damag After becoming aware of'he preliminary determination for the root cause of the stuck fuel assemblies.
the inspectors asked the licensee's staff about previous problems associated with the upper guide structure.
During a meeting on July 27.
1983.
the licensee's architectural/engineering contractor.
Bechtel Power Corporation.
recommended that the licensee develop guides for the upper guide structure lift rig.
Subsequently.
during a meeting on October 28.
1983.
the licensee and its contractor discussed developing visual alignment aids to ensure proper alignment of the upper guide structure lifting upper alignment bushings.
and a modification to the nuclear steam system supplier.
Combustion Engineering.
upper guide structure laydown brackets.
These efforts were to prevent possible damage to the guide tubes.
Subsequently.
the licensee installed I/4 inch diameter alignment rods (antenne),
designed for centering the upper guide bushings of the upper guide structure when lowering onto the laydown pads.
The guide bushings normally fit over the reactor vessel alignment pins assembly.
which were over 6 inches in diameter.
during upper guide structure installation into the reactor vessel.
During the current outage.
measurements between the alignment rods
'nd the guide bushings revealed that when the upper guide structure was resting on the laydown pad assemblies.
one alignment rod was 1 inch above the guide bushing and the other was 1/2 inch below the other bushing.
Since the tapered riser section of the laydown pad assemblies rise to a level 9-1/2 inches above the horizontal support of the laydown pad assemblies and guide tubes extend down about 10 1/2 inches below the core alignment plate.
there was potential for the guide tubes to impact the laydown pad assemblies when the upper guide structure was lowered to within about 20 inches above the horizontal support of the laydown pad assemblies.
At the time of the inspection.
the guidance for removing and replacing the upper guide structure in the reactor vessel was contained in Procedure 31NT-9RC33.
"Reactor Vessel Upper Guide Structure Removal and Installation." Revision 9.
Revision 9 was in effect when the upper guide structure was last removed from the reactor vessel.
The licensee provided Revision 5. which was in effect when the upper guide structure was removed and installed for Refueling Outages 2-5.
The licensee also provided a copy of the summaries of the changes for Revisions 6.
7.
and 8.
The changes in Revision 6.
~
and 8 were not relevant to the controls over moving and storing the upper guide structure.
The inspectors reviewed Revision 5 and determined that personnel removing or installing the upper guide structure would not be alerted by the procedure to the potential for damage.
There were no cautions.
notes.
or procedure steps that would heighten the awareness of handlers to exercise care not to inadvertently impact the upper guide structure on the lay down pad assemblies.
Also. the revision summaries subsequent to Revision 5.
and Revision 9 did not indicate that procedure enhancements had been implemented to address the risk of upper guide structure damag (
l
-22-The licensee's updated final safety analysis report references Combustion Engineering Standard Safety Analysis Report.
Section 3.9.5. 12 of the Combustion Engineering Standard Safety Analysis Report states that the design basis of the upper guide structure is to align and laterally support the upper end of fuel assemblies within the vessel.
This is accomplished by the upper guide structure suspended guide tubes.
which were designed to engage the guide posts on the fuel assemblies and. therefore.
position the upper end of fuel assemblies within the core.
The inspectors considered the failure to translate into the procedure adequate measures to assure the design basis of the upper guide structure was maintained.
to be a violation of Criterion III of Appendix 8 to
CFR Part 50.
This item will be tracked as Violation 529/9606-01.
The inspectors concluded that the licensee's root cause investigation team conducted their activities in accordance with the prescribed procedures and management expectations, Further.
the team leader promoted the team's effectiveness toward identifying the root cause of the stuck fuel assembly.
2.8 Com leted Interim Corrective Actions The licensee made repairs to the upper guide structure and modified a
procedure to enable the replacement of the upper guide structure into the reactor vessel and to return the unit to power operations.
Management indicated that a hardware design change would be developed and implemented.
2.8. 1 Upper Guide Structure Repair After extensive engineering evaluation.
the licensee determined that correct dimension. configuration.
and orientation of the guide tubes could be restored using the underwater electric discharge machining process that was used to cut the relief slots on the stuck fuel assembly's lower end fitting legs.
The inspectors held discussions with the licensee contractor representatives.
were shown the repair equipment.
and reviewed the repair procedures and drawings.
The same contractor that provided the damage assessment measurements performed the repairs on the upper guide structure by using robotically positioned and controlled equipment.
k Before and during the upper guide structure repair process.
the repair crew experienced numerous problems with the robot. including a need to redesign and replace the robot umbilical,'his activity was successfully accomplished.
The initial repair effort to the upper guide structure was to restore the minimum diameter to one guide tube over the Fuel Assembly A07 location.
The second operation was to machine cuts in two guide tubes over Fuel Assembly A08 location.
After the cuts were made in the guide tubes over the Fuel Assembly A08 location.
the contractor inspected the electric discharge machine electrode and expressed doubt about the adequacy of'he initial cuts to restore the desired tube dimension.
The decision was made to continue the effort to restore one guide tube over the Fuel Assembly T10 location.
After comparison with the electrode used to machine the guide tubes over the Fuel
-23-Assembly A07 location and addition'al visual inspection. it was decided to remachine both guide tubes over the Fuel, Assembly A08 location.
At the conclusion of all machining.
the robotic equipment was reconfigured for gauging.
Gauging was performed to assure adequate dimension. configuration.
and orientation of the guide tubes that had been restored.
On the basis of the engineering evaluation of the measuring results, the licensee decided that the upper guide structure was ready for installation into the reactor vessel.
The inspectors observed that the licensee had good interface with the contractor and effective oversight of the contractor's activities during the repair phase of the recovery.
The inspectors concluded that the licensee's effort to assess and repair the damage to the upper guide structure was adequate to provide -reasonable assurance that the structure would perform its intended functions.
The electric discharge machine repairs on the upper guide structure were well planned and performed in a conservative and thorough manners with appropriate verification of machining.
where necessary.
2.8.2 Revised Procedure for Upper Guide Structure Installation The licensee took various actions to decrease the likelihood of damaging the upper guide structure during reinstallation into the reactor vessel.
The inspectors assessed the effectiveness of these actions.
Revision 10 to Procedure 31HT-9RC33 provided changes to assure that the upper guide structure would be lifted cleanly from the laydown pads, safely transported'nd lowered carefully into the reactor vessel.
Listed below are the specific elements of the revision.
~
Required underwater video cameras to monitor the upper guide structure
, liftoff from the laydown pad assemblies.
~
Required the polar crane to be centered over the upper guide structure and verified this by assuring that structure remained within the tapers of the laydown pad assemblies when lifted.
~
Required the installation activities be stopped and an inspection and engineering evaluation be conducted if the guide tubes contacted any reactor cavity support structures.
~
Forbid the installation in to the reactor vessel of a potentially damaged upper guide structure.
~
Forbid lateral movement of the upper guide structure until a specific height above lay down pad assemblies was achieved.
The inspectors discussed the changes with licensee representatives in order to understand how the changes would achieve the desired results.
The inspectors concluded that the likelihood of damage to the upper guide structure would be
e
-24-significantly reduced by implementing the procedure changes as discussed above.
2.8.3 Preparations and Prejob Brief for Upper Guide Structure Installation The inspectors attended two separate briefings prior to the actual reinstallation of the upper guide structure.
The briefings were conducted early during the second shift on April 23.
1996.
The first briefing was given only to the personnel who were to be involved in actual movement of the upper guide structure.
The second briefing was conducted about three hours later and was an integrated briefing provided to maintenance.
radiological protection.
operations.
engineering'nd quality assurance personnel.
The two briefings provided guidance on the following elements.
Required plant conditions.
Participant tasks.
Sequence of events.
Acceptance/success criteria.
Expected results.
Criteria and methods for stopping the evolution.
Identification of condition or behavior for stopping the evolutions Problems to anticipate, Action in the event of personnel uncertainty.
and Communications.
Both briefings involved significant interaction between participants.
Participants asked several questions and presenters provided clear logical responses.
Radiological protection personnel provided detailed radiological conditions and requirements to be followed. including identifying low dose areas.
After one briefing. personnel assigned the task of operating the control element self-latching meehan'isms were required to use a mock up mechanism to practice performing the task.
Prior to moving the upper guide structure.
the inspectors observed sufficient personnel in the containment building to accomplish all involved tasks safely.
There were sufficient supervisory and management personnel on-hand to make on-the-spot'ecisions regarding the upper guide structure movement and evaluation in the event of an unexpected mishap.
This included an engineering supervisors'
quality assurance supervisors'nd the maintenance support department leader.
The inspectors observed the lift of the upper guide structure from the storage area and placement into the reactor vessel.
During the lift. transport.
and placement'ppropriate maintenance supervision was present in containment.
The inspectors observed good radiation protection practices.
The inspectors concluded that the ',icensee proceeded in a controlled and cautious manner during the handling of the upper guide structure.
The inspectors concluded that the licensee's interim corrective actions were adequate.
The revised procedure provided a significant reduction in the
-25-likelihood of damage to upper guide structure during installation.
and the prejob briefings were exceptional in scope and depth of coverage.
The adequacy of the repairs to the guide tubes.
however.
could not be definitively assured until the next unit refueling.
Overalls the licensee's preparations to correctly install the upper guide structure were noteworthy.
2.9 P1anned Lon -Term Corrective Actions The inspectors asked a licensee representative if there were any previous occurrences of difficulty in removing fuel assemblies during core defueling.
The representative indicated there had been two other occurrences at the Palo Verde site.
However.
both of these occurrences resulted because of fuel assembly bowing.
There were no previous instances where fuel assemblies had stuck to the core support plate as had happened in this case.
On Hay 15, 1996.
the inspectors received a copy of the licensee's root-cause investigative report.
This report identified the root cause of the stuck fuel assembly to be the deformed upper guide structure improperly engaging Fuel Assembly A07 and crushing it.
The root cause of the deformed upper guide structure was identified as:
~
An inadequate design of.the alignment system for placing the upper guide structure into the storage pit.
The design did not assure positive alignment prior to the guide tubes being able to come into contact with the laydown pad assemblies; and
~
~
An incorrect assumption by personnel involved in charge of the upper guide structure lifting operations that as long as the alignment bushings were higher than the alignment rods'he upper guide structure would not be damaged by the laydown pad assemblies.
The management-approved investigative report contained the significant long-term corrective actions listed below:
Scope alternative methods to positively align the upper guide structure prior to its being able contact laydown pad assemblies and present the methods to the modification review team:
Design such an alignment system; Implement the design change in each unit:
Implement a method for determining the lower alignment plate minimum elevation prior to lateral movement of the upper guide structure:
Inspect the Units I and 3 guide tubes for damage.
Initiate any repairs necessary:
Analyze the data from al'1 units'pper guide structure inspections.
Evaluate differences in amount.
location.
and damage mechanisms:
(
C f
!
f
-26-
~
Include the event in the quarterly industry events training for fuel handling personnel.
appropriate engineering disciplines.
and nuclear fuel management personnel.
The due dates for completing the long-term corrective actions were sequenced to the units'lanned refueling outages and extended out to November 15.
1997.
The inspectors considered the proposed long-term corrective actions to be appropriate.
Inasmuch as the same procedures and hardware will be involved in the future handling 'and placement of the Unit 1 and 3 upper guide structures.
NRC will be following up the licensee's long-term corrective action during the next defuelings of those units.
This followup will be tracked as a part of the followup to the upper guide structure design violation discussed above.
2. 10 Generic As ects of the Stuck Fuel Assembl Event Based on information developed by the licensee.
industry notifications of the stuck fuel assembly were issued on March 29.
1996 (NUCLEAR NETWORK Plant Status 5136).
and April 2.
1996 ( Institute of Nuclear Power Operations Significant Event Notification 135).
Both articles were entitled
"Damaged Fuel Assembly Found During Core Defueling."
The NRC inspectors contacted representatives of the three pressurized water reactor nuclear steam supply system manufacturers (ABB-Combustion Engineering'nc.;
Inc.; Framatome Technologies, Inc. [formerly Babcock and Wilcox, Inc.j) and determined that the Palo Verde upper guide structure design was unique to the ABB-Combustion Engineering System 80 design.
The Palo Verde units are the only domestic plants of the System 80 design.
The inspectors learned that at least one other ABB-Combustion Engineering reactor (Palisades)
has an upper guide structure design that, while different from Palo Verde.
incorporates a guide pin arrangement that interfaces with the top of each fuel assembly.
The inspectors learned during discussion with the NRC Senior Resident Inspector at Palisades.
that a fuel assemoly had been "stuck" to the guide pins during each of the last three refueling outage lifts of the upper guide structure.
It appeared that other ABB-Combustion Engineering reactor upper guide structures vere essentially flat with a chamfered hole that interfaced with the fuel assemblies.
The lowermost portion of a Framatome (Babcock and Wilcox) upper guide structure consists of machined and chamfered blocks that are approximately 5 inches long. 1-1/2 inches wide.
and 2 inches thick. and are bolted to the bottom of the core alignment plate, The Westinghouse equivalent upper guide structures are of two types:
one is a flat plate.
and one uses guide pins that are approximately 3/4 to 1 inch long.
The manufacturers'epresentatives were not aware of any of their customers having had damage to a fuel assembly as a result of damage to their equivalent upper guide structure e
-27-
ONSITE FOLLOWUP OF PREVIOUS INSPECTION FINDINGS AND LICENSEE EVENT REPORTS (92902)
3. 1 0 en Violation 528 529 530/94035-01:
Failure to Initiate a Corrective Action Document When Restrictions Were Im lemented to Assure 0 eration Within the Desi n Basis The licensee's reactor engineering staff identified a discrepancy during review of a procedure.
Engineering recognized that the fuel vendor's guidance to maintain boron restrictions or core protection calculator operable was insufficient.
The licensee's staff questioned that both the boron restriction and the core protection calculators were necessary.
The issue was referred to nuclear fuel management personnel.
without initiating a condition report/disposition request.
The NRC accepted a licensee response to the Notice of Violation that committed to:
(I) initiate a corrective action document identifying a knowledge weakness of the corrective action system by fuel management personnel; (2)
require nuclear assurance to brief fuel management personnel on identi fying the need to initiate a corrective action document; (3) provide training on the condition report/disposition request procedure to management:
and, (4) require individual managers to brief their own personnel.
Information available to the inspectors for closing this item did not contain documentation supporting the completion those actions identified in Nos.
(3)
and (4) above.
This item remains open pending documented completion of committed actions in Nos.
(3) and (4).
3.2 0 en Licensee Event Re ort 530/94-08:
Inadvertent Cross Wi rin of Standard and Mini Incore Instruments, The licensee identified personnel error as the cause of miswiring 3 of 5 mini incore instruments.
The output from these 3 mini incore instruments were inappropriately connected to the core operating limit supervisory system instead of to the standard incore instrument output.
This caused less conservative planar peaking factors to be utilized by the core protection calculators.
The licensee further identified as contributors to the error.
an inadequate prejob brief and insufficient detail in the work documents used to e
erform the work.
Also identified as lesser degree contributors.
were poor ighting and high radiation levels in the work area.
Information provided to the inspectors for closing this item contained documentation verifying the completion of actions to (1) remove the temporary modification that installed the mini incore instruments and.
(2) evaluate the effectiveness of programmatic configuration controls when temporarily modified structures.
systems.
or components are disassembled or worked on.
The information available to the inspectors did not document any corrective action that directly addressed the personnel error.
In additions there were no documented plans to address the significant contributors of an inadequate
4i I
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-28-prejob briefing and the insufficient work instructions.
This item remains open pending the licensee's identification of corrective action to address the above issues.
REVIEW OF UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR)
COHHITHENTS A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that compares plant practices.
procedures and/or parameters to the UFSAR descriptions.
While performing inspections discussed in this report.
the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected.
The following inconsistency was noted between the wording of the UFSAR and the plant practices.
procedures and/or parameters observed by the inspectors.
As discussed in Section 2.7.2 above.
the licensee's procedure for the handling of the upper guide structure was not adequate to ensure that the upper guide structure's design basis.
as given in the UFSAR.
was maintaine (
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ATTACHMENT
PERSONS CONTACTED 1.1 Licensee Personnel
- J. Bailey, Vice President.
Nuclear Engineering
- H. Bieling. Manager.
- R. Buzard, Primary Plant Event Investigator.
Nuclear Assurance
- T. Cannon, Department Leader.
Engineering
- D. Fan, Section Leader.
System Engineering
¹*B. Grabo.
Section Leader, Nuclear Regulatory Affairs
- R. Hazelwood.
Senior Engineer, Nuclear Regulatory Affairs
¹ J. Hesser.'Director.
Nuclear Engineering
- W. Ide, Director. Operations
¹*A. Krainik, Department Leader.
Nuclear Regulatory Affairs
- D. Mauldin. Director, Maintenance
~J.
McDonald, Director. Communications
- R. Nunez.
Manager.
Training
- G. Overbeck.
Vice President.
Nuclear Support
- M. Powell, Department Leader.
Design Engineering
- M. Radspinner.
Section Leader, Design Engineering
- M. Reid. Section Leader, Nuclear Fuel Management
- M. Shea.
Director, Radiation Protection
- D. Smith. Director. Outage/Scheduling
- W. Stewart.
Executive Vice President 1.2 Arizona State
~A. Godwin. Director. Arizona Radiation Regulatory Agency 1.3 NRC Personnel
- K. Brockman.
Deputy Director. Division of Reactor Safety
- J. Kramer.
Palo Verde Resident Inspector
'n addition to the personnel listed above.
the inspectors contacted other personnel during this inspection period.
- Denotes personnel that attended the exit meeting on May 1.
1996.
¹Denotes personnel that participated in the telephonic exit meeting on May 22.
1996.
EXIT MEETING An exit meeting was conducted on May 1.
1996.
Following the exit meeting.
in office review was performed on:
(1) the licensee's and the NRC staff's dose calculations of potential exposure to licensee personnel had fuel rod perforation occurred in Fuel Assembly A07 during the first attempt to lift the assembly.
(2) the licensee's root cause report.
(3) various interim corrective actions.
(4) whether other power reactor licensees might have upper guide structures'susceptible to damage as discussed in this report.
and (5) the
If a
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~ AS-2-significance of the potential design violation.
Subsequently.
a telephonic exit meeting was conducted on Hay 22.
1996.
During these meetings.
the inspector reviewed the scope and findings of the report.
The licensee did not express a position on the inspection findings documented in this report.
During the onsite exit. the licensee's representative expressed concern about the safety significance that NRC might attribute to the potential design violation.
The licensee did not identify as proprietary any information provided to. or reviewed by. the inspector f i