View source for B12563, Forwards Summary of Consolidation Process Evaluation,Per NRC Request.Util Reviewed Hot Demonstration Consolidation Process & Determined That Project Safe & Technically Acceptable

{{Adams
| number = ML20237J005
| issue date = 08/11/1987
| title = Forwards Summary of Consolidation Process Evaluation,Per NRC Request.Util Reviewed Hot Demonstration Consolidation Process & Determined That Project Safe & Technically Acceptable
| author name = Mroczka E, Werner R
| author affiliation = NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
| addressee name = 
| addressee affiliation = NRC OFFICE OF ADMINISTRATION &amp; RESOURCES MANAGEMENT (ARM)
| docket = 05000336
| license number = 
| contact person = 
| document report number = B12563, TAC-65274, NUDOCS 8708170457
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 42
| project = TAC:65274
| stage = Other
}}

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0 l                                                         NORTHEAET               UTILITIES cenerei Ott ces . seiden street, Berlin, Connecticut l          urm..mmm ucmc c"^*                                          P.O. BOX 270
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HARTFORD. CONNECTICUT 06141-0270                   J l
k    k   l NNNE c5^$                                                   (203) 66s-5000 August 11, 1987 Docket No. 50-336 B12563 Re: 10CFR50.59 l
U.S. Nuclear Regulatory Commission                                                                                     )
Attn: Document Control Desk l                                                                 Washington, D.C. 20555                                                                                                ,
i Gentlemen:
Millstone Nuclear Power Station, Unit No. 2 Fuel Consolidation Program Section 50.59 Evaluation of the Consolidation Process                                               i On June 2, 1987,(1) the NRC issued an amendment to the Millstone Unit No. 2 operating license to allow storage of consolidated fuel in the spent fuel pool.                                     _,
Although this amendment limits spent fuel consolidation to a hot demonstration                                       l program during which ten spent fuel assemblies will be consolidated into five consolidated storage boxes, the NRC Staff is c9ntinuing to review Northeast Nuclear Energy Company's (NNECO) May 21,1986t2) amendment application to allow full scale consolidation of spent fuel.
The Millstone Unit No. 2 fuel consolidation program hot demonstration will be conducted under the provisions of 10CFR50.59. The demonstration will involve a coordinated disassembly of two spent fuel assemblies and a subsequent systematic reconfiguration and repackaging of the 352 fuel rods into a consolidated spent fuel storage box.             The non-fuel bearing components (skeletons) will be volume reduced and packaged into a waste container. The scope cf the demonstration is to consolidate a total of ten spent fuel assemblies into five consolidated storage boxes for long-term storage in the spent fuel pool.
NNECO has reviewed the hot demonstration consolidation process generically and has determined that the project is safe and technically acceptable. Also, as summarized below, NNECO has reviewed the process against the criteria of                                              _
10 CFR 50.59 and determined that no unreviewed safety question is involved.
The 50.59 evaluation is also applicable to the full scale consolidation campaign.                                       i The process and associated risks and accident analyses are essentially the same                                         j regardless of the scope of consolidation.                                                                              J (1)    D. H. Jaffe letter to E. 3. Mroczka, dated June 2,1987, " Amendment No,117 to Facility Operating License No. DPR-65, Millstone Nuclear Power Station, Unit No. 2."
(2)    3. F. Opeka letter to A. C. Thadani, dated May 21,1986, " Millstone Nuclear Power Station, - Unit No. 2, Proposed Change to Technical                                    :
9 0(
Specifications, Storage of Consolidated Spent Fuel."
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8708170457 870811 DR        ADOCK O           g6

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U'S. Nuclear Regulatory Commission B12563/Page 2                                                                        ,
August 11,1987                                                                      i l'
NNECO's evaluation has specifically considered the nuclear criticality, thermal-    !
hydraulic, radiological, civil, mechanical, materials, and seismic / structural aspects of the consolidation process.. This included evaluation of:
o the nuclear criticality and thermal-hydraulic considerations related to    j the consolidation process; o the equipment and components of the consolidation system process; o the hoist assembly and consolidation workstation; j                  o potential offsite dose consequences of postulated accidents related to l                     the process; o the mechanical, material, seismic / structural aspects of the temporary (3 x 3) spent fuel storage rack design; and 1
o postulated accidents involving the use of the existing spent fuel pool platform crane for the purpose of moving consolidated fuel storage boxes from the cask laydown area to the spent fuel storage pool.
The design of the consolidated fuel storage boxes has been previously evaluated      )
in conjunction with NNECO's proposed amendment to allow storage of                    '
i          consolidated fuel.         Also in that context, NNECO has evaluated the             !
: j.         nuclear / criticality and thermal-hydraulic' considerations related to storage of consolidated fuel.
As per your request, a summary of NNECO's review and 10CFR50.59 evaluation of the consolidation process is included as Attachment 1. We hope you will find      ;
this information satisfactory and we remain available to answer any questions        !
you may have.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Ax\ cn  ~
l                                                  E. J. Mr6czka           d Senior Vice President 3 .h Q n Af     .es.- .~
                                                ,By: R. P. Werner Vice President cc:    W. T. Russell, Region I Administrator D. H. Jaffe, NRC Project Manager, Millstone Unit No. 2 T. Rebelowski, Resident Inspector, Millstone Unit Nos. I and 2
                                                                                              -]

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l Docket No. 50-336 B12563 i
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l l                       Attachment 1                                                                                                             l l                                                                                                                                                1 1
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{. consolidation Process Evaluation i
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August,1987

u U. S. Nuclear Regulatory Commission                                                                                                1 B12563/ Attachment 1                                                                                                               J I
Page1 1.0   Consolidation Process l
1.1    Process Overview                                                                                                             ]
1 NNECO has developed a consolidation system which permits the                                                                   j I
l coordinated disassembly of spent ' fuel assemblies and subsequent                                                              '
t repackaging of the rods into dense close packed arrays. For this program, the actual' in-plant work will be performed in the Millstone Unit No. 2                 .
i'j spent fuel pool cask laydown area. Here, the respective work stations, fuel support components, hoisting equipment, filtration system, and various handling tools, whleh comprise the consolidation system, are installed.
Also located within this general area is the main control console from which operations will be maintained and controlled. A temporary storage rack, which is serviced by both the fuel handling machine and the i
consolidation system holst, holds both intact and consolidated assemblies.
1 I-Under NNECO's process, fuel assemblies to be consolidated, having a                                                           ,
i minimum of 85% burnup and 5 years subsequent residence time in the pool, are deposited in the temporary storage rack by the spent fuel handling
!                   machine. Consolidation takes place in the work station frame, which supports seven individual work stations. A traversing carriage on the work station upper plate aligns fuel manipulating equipment accurately with the individual stations.
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                                                          - U. S. Nuclear Regulatory Commission                                                                 !
B12563/ Attachment 1 Page 2 -
                                                                                                                                                                )
i The fuel is disassembled in the first station by' cutting off the upper end     j fitting and then removing fuel rods one row at a time by means of a multiple rod pulling tool '(MRPT).      Rod pulling forces are maintained between preset limits. Fuel rods are then deposited in an interim transfer     j l
canister (lTC) at the next work station. The X-Y positioning table locates      1 the MRPT at any point above the fuel disassembly. station, ITC- station,     .)
i damaged fuel rod station, or the recovery (separation) station. .The           ;
positioning table is hydraulically actuated under remote control and is equipped with linear variable differential transformers (LVDT)'to permit precise location.
l l                                                                               The ITC has channels which guide the rods into a close packed triangular l
array at the bottom. (Damaged fuel rods, if encountered, are deposited in l                                                                               the damaged fuel rod storage station'af ter being separated from the row in     j l
!                                                                               the separation / recovery station). The fuel rods in the ITC are normally I
transferred by gravity into a consolidated fuel storage box located in the      I next work station. Descent of the rods is controlled by a telescoping cylinder through the bottom of the box. A rod transfer tool can also be used to assist in this operation, .if required. The close packed triangular rod array results in a compaction ratio of 2:1.         Lockable covers are installed on the filled consolidation fuel boxes before they are moved for storage in the spent fuel pool.
Fuel assembly end fittings are placed in storage boxes at the end fitting storage station. Grid cages and control guide tubes are compacted by hydraulic cylinders in the compaction station.

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U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 3 L
I A filtration' system is connected to the fuel disassembly,ITC,' transfer and-l                  compactor work stations to collect and filter out radioactive particulate       i
!                                                                                                   t generated by operations performed at these stations. In addition, the            ;
MRPT is' enclosed by a shroud assembly which provides guidance and alignment, and vents released gases to the plant gas handling system that j                 may result from a broken rod,if encountered. ,
L                                                                                                   j The consolidation system is controlled from a panel at a work platform on the hoist assembly. Control logic is programmable to allow for changes as 1
experience is gained. A TV system is provided for remote vieuing of all consolidation operations.                                                        '
l 1.2   Proposed Plant Design Changes i
The fuel consolidation process is planned to be conducted under the provisions of 10CFR50.59. The permanent changes to' the plant resulting from the fuel consolidation program are as follows:
o        Removal of two lighting support brackets: within the cask laydown area there are two lighting support brackets that must be removed in order to     install  the consolidation workstation without any interference.
o        Installation of workstation wall brackets:    the installation of the workstation requires that two stabilizer brackets be welded to the liner surface at approximately 170 inches from the pool floor, a

                                                                                                                                                      'l.
U. S. Nuclear Regulatory Commission B12M3/ Attachment 1 Page 4 o     Installation    of    the   rail   system:     the               operation of                          the t
bridge / trolley / hoist equipment above the cask laydown. area requires l                             that a pair of rails be installed.
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l                                                                                                                                                       1 l                       o     Bracket modification for level switch and temperature element: On                                                         I I
the intermediate wall of the cask laydown area, there exists a series of brackets that support the level switch and temperature elements i
in the spent fuel pool. In order to install the rail on the south wall                                                    i j                             without interference, the support bracket must be modified by .
l repositioning the bolted connections.
                                                                                                                                                        ?
o      Modification of the spent fuel pool handling machine " sweep" plates:
in order to install the rail on the north side of the cask laydown area without interference, the north " sweep" plates on the spent fuel pool fuel handling machine must be partially coped out.
i o     Installation of the tool support base plates: The operation of the                                                         )
I consolidation equipment requires that the individual remote handling                                                       1 I
tools and their storage brackets have a permanently mounted base                                                             j 1
plate for attachmes :. The tool support base plates will be welded to                                                       !
the embedded angle on the top periphery of the cask laydown area.
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                                                                                                                                                      ~'
o     Installation of the new spent fuel handling tool: a new spent fuel handling tool is required in order to handle the weight of the consolidated storage boxes in and over the spent fuel pool and to 1
i

U. S. Nuclear Regulatory Commission                                                      l B12563/ Attachment 1 Page 5
!                                                                                                    properly interface with the height'of the new spent fuel racks while placing the consolidated storage boxes into the rack cavity.
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                                                                                                                                                                                 '1 1
o    Requalification of the existing spent fuel pool platform crane: In           j order to handle the weight of the consolidated storage boxes in and.        j over the. spent fuel pool, the existing hoist must be requalified for a 4
l                                                                                                    limited number of consolidated storage box movements (hot                     l 1
demonstration).                                                              I l
l o    Consolidation of fuel assemblies into consolidated storage boxes: All required mechanical and nuclear design and analyses issociated with the consolidated storage box and spent fuel racks . have been i                                                                                                    submitted to the NRC in the May 21,.1986- license amendment l
application. The consolidated storage box configuration is integral to the design and analysis of the spent fuel racks and . the license amendment request for storage capacity expansion at Millstone Unit No.2. The spent fuel racks and the pool / building structure are analytically qualified (seismic / structural) for storage of consolidated fuel with a density of 2:1 in every storage location.
2.0   Prevention or Mitigation of Fuel Damage in Normal Consolidation Operations The principal features designed into the fuel consolidation equipment are described below. These are designed to prevent damage to fuel rods during

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                                                                             .U. S. Nuclear Regulatory Commission
                                                                            - B12563/ Attachment 1-Page 6 -
normal consolidation operations, or, in conjunction with operating procedurei, should limit the extent and effects of such damage should it occur. Fuel consolidation Hot Demonstration operating procedures will also. preclude the consolidation of fuel assemblies known in advance to contain one or more damaged fuel rods. . In conclusion, the design of the consolidation equipment. meets all' applicable criteria for normal and abnormal consolidation process operations, as described below.
2.1   General Features                                                             !
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l 2.1.1   Operator Contrcl l
All operations are under manual control of the operators from the    i control console, all key locations can be viewed by the TV system, and individual controls on the console are interlocked so as to      <
l minimize operator error. An emergency stop button on the console     i enables the operator to stop all operations at once should this      h 1
become necessary. A moderate target rate .of consolidating one      j s
fuel assembly per eight hour shif t is planned.                      l 2.1.2   Fail-Safe Design                                                     '
I (a)    Electrical Failure                                            l 1
1 Equipment is designed to stop in a safe mode on electrical failure. For example, the hoist assembly stops and the

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          .U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 7 brakes are automatically set; the X-Y table control valves go to the locked position; the transfer piston control valve will go to the locked position; and the MRPT grippers are spring loaded to fail closed.                                                     !
(b)   Air Pressure Failure i
The grippers of the MRPT are spring loaded to close on loss of air pressure, maintaining a grip on fuel rods. The control console includes a visual alarm of low' or high air pressure.
l (c)    Failure of Water System i'
This system is the source of propulsion for the X-Y table.
With loss of pressure, the X-Y table will not move. The l
i control console includes an indicator light identifying a low                   'l I
or high pressure condition.                                                        l 1
2.1.3  Control of Crud, Particulate and Radioactive Gases i
The filter system consists of two trains of filtration. Each train     .
consists of a pump, strainer and two filters.         The trains are operated in parallel and are manifolded to four work stations to                           ;
provide positive downward flow through any of the four stations.-
The pump and filter units are sized to ensure that there is i
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U. S. Nuclear Regulatory Commission-B12563/A ttachment 1 l                    Page 8 sufficient velocity at the MRPT and work station interface to f                                  entrain any released radioactive crud before it can disperse within I.
the cask laydowr, area.      The filters are capable of removing particles as small as 2 microns, corresponding to a predicted crud removal of greater than 99%. The pump / filter suction manifold has branches connecting to the fuel disassembly station, ITC l                                 staCon, transfer station, and compactor station. Each of these 1
branches contains an isolation valve and a flow sensor. The valves are used to balance the flow between stations, or isolate unused stations, in order to maximize the crud entrainment velocities within the stations in use. The flow meters inform the operator of                                                                  )
the flow balance within the circuits. The meters also ' provide a low-flow alarm when the flow falls below a pre-set value, indicating filter expenditure.
The MRPT is enclosed in a shroud which is designed to capture any                                                                   I gases released during rod removal that may result from a broken                                                                     I rod. This shroud is vented to the Millstone Unit No. 2 Fuel Handling Ventilation System. The filter system provides a positive downward flow through the workstation shrouds to collect crud and possible loose particles, and may also entrain gases if a gas release were to occur. For this reason, the filter system also has vents connected to the Fuel Handling Ventilation System.

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U. S. Nuclear Regulatory Commission l          B12563/ Attachment 1 l          Page 9 2.2   Specific Fuel Rod Handling Safeguards l
I 2.2.1    Hoisting (a)     Cable Failure The hoist assembly is designed in accordance with industry specifications and has a design capacity of 6000 lbs, with 1
the load capacity limited to approximately 3500 lbs. This                                      ;
I results in'a working cable safety factor of approximately                                      l 10:1 based on the ultimate tensile strength. The potential consequence of a cable failure is a dropped fuel assembly or consolidated fuel storage box accident. These accidents are discussed in Section 8.0.
I (b)     Brake Failure The hoist assembly has two independent brakes; one a large                                     j e
l mechanical load brake, the other a rectified dc disk type                                      j electric motor brake (when power is off the brake is on).
t (c)     Control Failure l
The operator can turn off the power to the hoist system or                                     ,
use the emergency stop button on the control console to stop the entire operation immediately.
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                        . B12563/ Attachment 1                                                                   i Page 10                                                                                  I l                               2.2.2   Fuel Rod Grappling                                                     -l j                                                                                                                l (a)   Incorrect Lowering of MRPT Head and Ramming of Fuel 1
Rod                                                                 I l
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The MRPT head includes an LVDT system for each of the               l grippers. The LVDT provides information to control logic of the penetration of a rod within the pulling grippers. The.         1
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MRPT control console logic includes maximum and minimum             i
,                                             rod insertion limits to prevent accidental ramming of a rod, I
or gripping of a rod with insufficient gripping . area, respectively. The hoist assembly is electronically locked-l                                             out upon reaching the maximum rod insertion limits to l                                             prevent further lowering of the MRPT.
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i (b)    Misalignment of MRPT with Fuel Rods or ITC o     At the fuel disassembly station - the operating procedure will specify to the operator the minimum load requirement on the hook. If the MRPT is being lowered and the gripper lands on top of a fuel rod, due
!                                                   either to misalignment or a closed gripper, the hoist descent will be interrupted when the minimum hoist j
load is reached, thereby preventing overloading of the fuel rod. Visual inspection with the camera system l
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p U. S.- Nuclear Regulatory Commission                                                                                                                            ,
.. B12563/ Attachment 1 j      Page 11 will identify the problem. The MRPT control console'
                                                                                               ' has a bar graph indicator for each gripper assembly l
which enables the operator to verify visually that the l                                                                                                fuel rod has properly entered the grippers.-
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!                                                                                         o     At the ITC station - the top of the -ITC includes i                                                                                               several design features to mitigate a misalignment l
problem.      The top plate of the ITC has .50 inch diameter holes to accept .44 inch diameter fuel rods.
l                                                                                               In addition, these holes have a 600 lead-in beginning with a diameter of .64 inch. The lower end cap of the fuel rod also has a taper bcginning with a diameter of
                                                                                                .32 inch. The operating procedure will. also require.
that    the camera                          systems be used for visual verification of proper alignment between the fuel rods.
and the holes in the upper plate of the ITC.
(c)         Excessive Pulling Force on Fuel Rod The MRPT is designed to limit the pulling force on any fuel rod to a normal maximum of 50 lb which can be increased to 100 lb in the over-ride mode. Past experience and resultant stress values indicate that this relatively low loading will not damage fuel rods. - In addition, the MRPT . grips the-
                                                                                                                                                                         -l outside of the fuel rod, rather than the upper end cap.

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U. S. Nuclear Regulatory' Commission B12%3/ Attachment 1
           . Page 12                                                                                                   ,
I (d)     Accidental Fuel Rod Release                                                          j l                                                                                                                       l l                                                                                                                      1 The MRPT control console contains interlocks to prevent                         -l the operator from accidentally releasing a rod. To release .a                     )
rod, the MRPT head must be in the lowered position and the i
operator must actuate two switches simultaneously. This                             ]
                                                                                                                   .l l                                  requires the use of both hands and thus forces him to make a very deliberate effort to release a rod. The main control.                          j 1
console includes a programmable controller (PC) which oversees the entire fuel consolidation process. The PC logic                        q l
will not allow the release of a rod if the MRPT is not at an appropriate work station such as the disassembly, ITC, or                          j separation stations.                                                            .l (e)     Drif t of X-Y Table While Rods are in MRPT                                         ,
The X-Y table is propelled by rodless cylinders powered by a                       i hydraulic water system. The table will be locked in position 1
by closing the inlet and outlet ports of the cylinders, thus                         l l
hydraulically locking the table in position. This method was                         ;
chosen over a vented system or a pressurized locking system.
1 A vented system would allow the table to drift if a side force were introduced. A pressurized system was discarded because, if a pressurized line were to rupture or leak, the X-                      l i
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1 U. S. Nuclear Regulatory Commission                                                            I J
B12563/ Attachment 1 Page 13
                                                                                                                                     -4 Y table would be driven.          In the chosen l design (i.e.,
hydraulically locked but unpressurized), if one line were to l
rupture or leak, the table would remain in the locked i
position.
There are no obvious side forces introduced during the hoisting of the fuel rods. ' The weight of the MRPT shroud acts vertically on the X-Y table. The mast and pulling head are supported on the hoist assembly which locates the tool i
in close proximity to the centerline of a work station. Thus,        j minimal side forces, considered to be of similar rnagnitude j                                                                                                                                     l l
l                                                                to friction forces in the system, are introduced.
i                                                                                                                                     i l                                                                                                                                    .I-2.2.3     Fuel Disassembly (a)    Damage to Fuel Rods from Guide Tube Cutter                           l l
j TWdesign of the guide tube cutter precludes the possibility of cutting into a fdgi rod adjacent to the guide tube. The cutter operates on the inside diameter of the tube, in which it is centered by the cutter body. Radial feed of the cutter            j x
is set by ?a fixed internal cam surface and it can only penetrate slightly beyond the outside diameter of the guide tube.
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U. S. Nuclear Regulatory Commission B12563/ Attachment 1                                                                 .
        - Page 14 (b)    Stuck Fuel Rod If a stuck' fuel rod is encountered, all other fuel rods will first be removed. Then, using the guide tube cutter, the i
grid cages will be removed one by one until the stuck rod
                              . can be freed with the single rod removal tool.
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(c)     Fuel Rod Lef t in Grid Cage Assembly                           )
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The operating procedure will require accounting for all fuel rods. The camera system will also be required to be used to i                              verify that there are no fuel rods remaining in the grid cage assembly prior to transferring it to the compactor station.
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2.2.4   Fuel Rod Transfer                                                     1 (a)    Crossing of Fuel Rods in the ITC Crossing of fuel rods within the upper 10 feet of the ITC box cannot occur because each rod is in a totally enclosed channel fabricated within the grid structure of the ITC. The    l partition design and the loading pattern mitigate the i
occurrence of crossing of rods below the upper section.         !

U. S. Nuclear Regulatory Commission l
l B12563/ Attachment I Page 15 However, if crossing of rods occurs within the corrugated sub-assembly of the ITC, all of the remaining rods will not fit into the ITC. Damage to an adjacent rod is unlikely since further insertion of fuel rods will not be possible, resulting in slippage within the MRPT gripper and stopping of the hoist.
(b)   Misalignment Between-ITC and Consolidation Fuel Storage Box I
!                             The operating procedure will require visual verification of alignment between the interim transfer canister and the consolidated fuel storage box.                                I l
The ITC gate mechanism incorporates a framework with lead-ins to facilitate alignment.        If the boxes are misaligned, the ITC will not rest properly on the stabilizing structures and the problem will be obvious because the ITC cannot be attached to the stabilizing structure.
The rods cannot be released if the ITC and the consolidated fuel storage box are not aligned because the transfer piston will not lift the transferable floor of the ITC and            l consequently the gate mechanism cannot be opened.

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d U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 16                                                                                !
(c). The design of the ITC provides a water lubricated smooth           I straight line passage which is essentially vertical for each     ,
1 fuel rod. Hence, hang up during transfer of the rods into the consolidated fuel storage box is not expected to occur.            j However, the rod transfer tool may be . used - to assist transfer if hang up is suspected, and it will also serve to 1
verify that transfer has been completed.                         -
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(d)   Incomplete Loading of Consolidated Fuel Storage Box The criticality analysis of the st.orage of consolidated spent fuel in Region 11 of the spent fuel storage pool is based on       ,
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the consolidated fuel storage boxes containing a minimum         . i number of fuel rods from the two intact fuel assemblies.
The' consolidation operating procedure will require this to be .
verified during loading of the ITC. The use of spacer rods l                             may be required to permit a transfer.                              I l
2.2.5  Damaged Fuel Rods (a)    Visual Identification TV camera systems are provided for visual inspection of
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rods as they exit the fuel disassembly station for identifi-cation of any broken or damaged rods. The camera systems l
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1 U. S. Nuclear Regulatory Commission                                                                                   {
B12563/ Attachment 1 -                                                                                                i Page 17                                                                                                                 l l
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are mounted 900 apart on the lower work platform and have a pan and tilt capability.
(b)   Detection by Gauging i
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A fuel alignment bar is incorporated in the MRPT to l
l                                     maintain fuel rod alignment and also to act as a gauge of l
possible over-sized rods.                          A blister may or may not . be removed while passing rods through spacer grids and alignment bars. If removal of a blister occurs, fragments are captured in the filter system. A blistered fuel rod may not pass through the alignment bar, resulting in slippage of the rod within the' gripper and identification as a damaged rod.
(c)    Bent Rods                                                                                       ,
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While severe bends have not been observed in past experience with fuel reconstitution, inspection, etc., a                                            l l
severely bent fuel rod would be treated as a " damaged" rod and stored in a damaged fuel rod storage box. A curvature of a fuel rod will not affect handling operations due to the rod's flexibility.

                       'U. S. Nuclear Regulatory Commission B12%3/ Attachment 1 Page 18 (d)     Removal of Broken Rods Intact fuel rods will first be removed in the fuel disassembly -
station. The CEA guide ttbes of the grid cage assembly will then be cut above the broken pieces and that portion of the l                                            assembly removed. The protruding piece of fuel rod will be '
removed using the single rod pulling tool. This process can be repeated to recover all remaining broken pieces. Fuel peitet fragments, if any, will be- retained in the fuel assembly station or entrained in the filter system strainer.
(e)    Storage of Damaged Rods.
Broken or damaged fuel rods are stored separately from intact fuel. Each broken or damaged fuel rod is contained in a separate chamber within a damaged rod storage box.
3.0   Radiological Evaluation As part of the evaluation of the fuel consolidation process, NNECO examined the potential offsite dose consequences of the following postulated accidents associated with the consolidated fuel storage box and the temporary fuel storage rack: (1) an assembly drop on a consolidated fuel canister (CFC), (2) a drop of a CFC, and (3) a drop of a spent fuel

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U. S. Nuclear Regulatory Commission l          B12563/ Attachment I
!          Page 19 f                                                                                                           i l                                                                                                           !
shipping cask. All calculations assumed the CFCs had decayed 5 years.
NNECO concluded that' all these postulated accidents will not increase the 1
site boundary doses for Millstone Unit No. 2 fuel handling incidents or                  j 1
result in offsite radiation exposure exceeding the limits of 10 CFR 100, l
i Due to the significant decay time, the consequences of the drop of 'an assembly onto' a CFC were shown to be bounded by the FSAR assembly                      1 J
drop analysis. The drop of a CFC was shown to be bounded by the drop of a spent fuel shippi:-; cask. Finally, for the cask drop, it v:as conservatively             1 assumed that 782 fully loaded CFCs were destroyed.                 The limiting exclusion area boundary dose was calculated to be 160 millirem to the                    1 i
whole body. This is lower than the previously analyzed cask drop dose of                  j 241 millirem. (Thyroid doses are negligible due to the 5 year decay of iodine). Therefore, the proposed consolidation process would not increase the consequences of the accidents previously evaluated.
4.0     Hoist Assembly and Consolidation Workstation In support of the fuel consolidation program, NNECO performed a civil structural safety evaluation of placement of the following items on elevation 38'-6" of the Millstone Unit No. 2 Auxiliary Building in the area j               of the spent fuel pool cask laydown area and elevation (-)2 - 0" of the cask laydown area:
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o        6,000 pound load capacity gantry styled hoist assembly, o        40 pound ASCE rail tracks and bearing plates.
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U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 20 o      Wall brackets for lower' work platform.
o      Lower work platform.
l 4.1    Holst Assembly / Rails l
                                                                                                               'l l The gantry type hoist assembly is similar in design to the spent fuel           !
l l                                 handling machine. The assembly spans' the cask laydown area and runs on rails installed on the operating floor. The rails ruri beyond the cask     l 1
I                                                                                                                 l laydown area so .that the hoist assembly can move out of the way of the l
spent fuel handling machine when the latter is working over t e cask laydown area. Hold-down bars, bolted to the ends of the wheel frames, I
capture the rails to prevent derailment. Horizontal guide rollers at the same locations are also utilized to prevent derailment.
l The hoist assembly has a 6000 pound load capacity with a total vertical travel of 52 feet; 22 feet of which is above elevation 38'-6".          The estimated actual load to be lifted will be 2,600 pounds (wet) and              i 2,900 pounds (dry). This represents a factor of safety of two on actual load versus design load in addition to actual design factors of safety. Seismic analyses were performed on the gantry assembly to determine interface loads between the hoist assembly and the rails during a safe shutdown 1
earthquake. The calculations determined the existing reinforced concrete walls to be structurally adequate to support the loading. All stresses in the hoist assembly were found to be within normal code allowables and                )
therefore within acceptable limits.
x l

a U. S. Nuclear Regulatory Commission                                                        .I B12563/ Attachment 1 Page 21 The gantry hoist assembly wheels run on 40 pound ASCE rail tracks which              -)
i are positioned on the north wall and south wall of the cask laydown area.              j I
The rails are divided into permanent and removable track sections. Each                 j i
40 pound rail is welded to the stainless steel baseplate in order to provide a          j composite rigid section that is capable of resisting the uplift forces.
NNECO's calculations demonstrated that the total hold-down capacity by                )
the four wheels provides a factor of safety on actual uplift forces of l                                                                                                                     approximately 5 which assures that the gantry will remain intact with the             I j                                                                                                                     track system.                                                                        'l I
Rail stops are also provided for each set of gantry assembly wheels at the west and east ends of the rail system. The rail stops are designed for a horizontal stopping force of 2,000 pounds, which is two times the actual stopping force as shown in a test of the gantry assembly at the manufacturer's facility.
NNECO also analyzed the use of the existing spent fuel pool platform crane for the purpose of moving five consolidated fuel storage boxes from the cask laydown area to the spent fuel storage pool during the hot demonstration program.       The spent fuel platform crane is rated for 2,000 pounds with a 25 percent overload capacity for a total capacity of 2500 pounds. The weight of a consolidated fuel box of 2,600 pounds is greater than the design crane capacity by approximately 4 percent.
NNECO has reviewed the spent fuel pool platform crane design and testing documentation to insure the acceptability of the use of the crane for the movement of fuel consolidated fuel storage boxes.
h l

U.S. Nuclear Regulatory Commission.
B12563/ Attachment i Page 22 Prior to the movement of the consolidated fuel boxes, the crane vill be tested and inspected to further insure acceptability.
l 4.2  Consolidation Work Station The lower work platform is located in the eastern portion of the cask l
l                laydown area. The platform is supported by six vertical columns with adjustable feet which stand on the cask laydown area floor. The platform is located and stabilized by two vertical guide pins projecting from the underside of the upper deck plate framing. The pins engage in holes in l                brackets which are attached to the pool wall. One of the holes is slotted to l
l                allow for differential expansion of the platform and the wall.               j l
f l                The floor of the laydown area is composed of a 5 foot thick reinforced 1
concrete slab on structural fill which in turn rests on bedrock.        The laydown area is designed to store a 100 ton spent fuel cask and has been analyzed for a 100 ton cask drop.      Since the vertical load for the work platform is 20,010 pounds, which is considerably less than a 100 ton cask, the floor of the cask laydown area is structurally adequate to support the load of the work platform.
Seismic analyses were also performed to determine bterface loads between the lower work platform and the pool wall and floor during a safe shutdown earthquake and to verify that the stabilizing guide pins will

           .                                                                                                     q U. S. Nuclear Regulatory Commission B12563/ Attachment I                        <
i Page 23 l
I remain intact, ensuring that the platform will not rock on its supports or fall over onto adjacent equipment. .The embedded wall angle / liner plate has been determined to be structurally adequate to support the imposed                          i loading without damaging the liner.
t 5.0   The 3 x 3 Temporary Spent Fuel Storage Rack i
The 3 x 3 temporary storage rack is a stainless steel monolithic honeycomb structure with square fuel storage locations. Storage locations are formed                      j 1
l                  by welding boxes, panels and angles at the seams. Semicircular passages at the bottom of every cell will allow cooling water to flow to all cells.
Reinforcing plates at the upper peripheral edges provide the required strength for handling.      Stainless steel bars inserted horizontally into                      l rectangular slots in the lower regions of the rack support the fuel assemblies or consolidated fuel storage boxes.
The rack is supported by a base plate welded to the bottom of the rack, l
which provides a low pressure bearing surface. Rack leveling, if required,                       I is performed in the field by welding shims to the bottom of the base plate.
An additional non-integral bearing plate for final placement of the rack will be provided during the Hot Demonstration to eliminate direct bearing of the welded shims on the pool liner surface,if required.
The 3 x 3 temporary spent fuel storage rack has been designed to meet the requirements of the Nuclear Regulatory Commission's position paper
                                                                                              .- -_____-__ _ - -_a

1 1
        .                                                                                                                                j l
U.' S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 24
                                                        " Review and Acceptance of Spent Fuel Storage and Handling Applications,"
dated January 18, 1979. Detailed structural and seismic analyses of the         .l l
rack have been performed to verify the adequacy of the design to                  l withstand the loadings encountered during normal operation, the severe and extreme environmental conditions of the Operating Basis Earthquake (OBE), Safe Shutdown Earthquake (SSE), and the abnormal loading conditions of an accidental fuel assembly and consolidated box drop events.      1 The rack is free standing (i.e., it is free to tip or slide horizontally-on the   j l                                                      pool floor) and is designed to assure rack structural integrity while at the i                                                      same time keeping the fuel in a safe, coolable, subcritical configuration          ,
i during all loading conditions.
Analyses have been performed to determine the maximum sliding and/or tipping displacement of the temporary storage rack during a seismic event.
The sliding analyses were performed for a fully loaded and an empty rack.
It was determined that the rack would not slide in either case. The tipping analyses were performed for a fully loaded, partially loaded and empty rack. It was found that the rack would remain stable and not overturn or interfere with either adjacent equipment or pool walls, provided that there is a minimum initial clearance of two inches around the rack.
6.0                         Nuclear / Criticality Considerations 6.1       The nuclear / criticality considerations related to the storage of consolidated fuel have been previously evaluated in conjunction

p                                                                                                                        q
        .                                                                                                                   q o
U. S. Nuclear Regulatory Commission B12563/ Attachment i Page 23 with NNECO's proposed amendment ' to allow storage of                                               ;
I consolidated fuel.       Detailed - below are      nuclear / criticality                          :q considerations specifically associated with the consolidation
                                                                                                                           'I process. The effective multiplication factor was evaluated for                                      j 1
these conditions for fuel which 'is acceptable for insertion in                                     j Region II as shown in the Millstone Unit No. 2 Technical                                            );
l Specification Figure 3.9.3, and was found to be less than 0.95.                                     l 1
1
* An isolated intact fuel assembly, assumed to be surrounded by                                        J unborated water at room temperature.
l
* An isolated consolidated fuel storage box containing the fuel rods from two intact fuel assemblies, assumed to be surrounded by l
l                       unborated water at room temperature.                                                              -!
                      *The 3X3 temporary storage rack containing nine intact fuel assemblies.                                                                                          ;
                      *The fuel disassembly station containing an initially intact assembly from which the fuel rods are extracted.
                      *The ITC station containing the fuel rods extracted from two intact assemblies in the fuel disassembly station.
h
____._._____.____________.w

U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 26
                                                                                                       *The damaged fuel rod station containing a storage box with fixed stainless steel tubes containing up to 196 damaged rods.
                                                                                                       *The fuel rod transfer station.
                                                                                                       *The entire consolidation work platform.
                                                                                                                                                                              .I 7.0     Thermal Hydraulic Considerations 7.1     The thermal hydraulic considerations associated with the storage of consolidated fuel have been previously evaluated in conjunction j                                                                                                       with NNECO's proposed amendment to allow storage of f                                                                                                                                                                               4 i                                                                                                       consolidated fuel.      Section 7.2 below addresses the thermal l
I                                                                                                       hydraulic considerations associated with the consolidation process, t
i 1
7.2     Fuel bundles that are to be consolidated must meet the two-fold I
criteria of having a burnup of 85 percent and a decay period in the      I spent fuel pool at least five (5) years.      During consolidation, a maximum number of ten fuel bundles will be moved into the cask laydown area at one time. The calculated decay heat addition to the cask laydown area from the ten bundles is expected to be 1
approximately 30,000 Btu /hr. Heat addition to the cask laydown area during consolidation from miscellaneous equipment including 1
the filter system and lights is expected to bring the total heat addition to the cask laydown area to approximately 145,000 Btu /hr.
u_____   _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . - . - - . _ _ _ _ _ . . _ . _ _ _ - - _ - - _ _                    - . - .   -.

U. S. Nuclear Regulatory Commission B12563/?.itachment 1 Page 27 i
During fuel consolidation, the gate separating the cask laydown area and the spent fuel pool will be installed, and the spent fuel pool cooling to the cask laydown area will be secured. The heat addition to the cask laydown area from decay heat and equipment will increase the temperature in the cask laydown area at a rate of                 1 l
J                                                                                                                       approximately 18 degrees per day. Therefore, during an eight-hour period, the water temperature in the cask laydown area is expected to increase only about six degrees. Because the majority of the addition of heat into the cask laydown area during consolidation is                 j from the auxiliary equipment, the water temperature in the cask laydown area will be monitored and evaluated by procedure during                     j l
intervals when the gate to the cask laydown area is installed. Fuel                  j i
1 consolidation may be performed continuously when the gate to the                       '
cask laydown area is installed and if the water temperature in the cask laydown area does not exceed 1200F. As auxiliary equipment 4
is intended for operation only when fuel consolidation is being                      4 performed, heat addition to the cask laydown area when consolidation is not being performed will be limited to decay heat addition from the spent fuel cells.       It should be noted that estimates for the amount of heat added to the spentfuel pool is conservative. The effects of heat transfer through the walls of the s
spent fuel pool, heat loss to ambient through system piping and                      I components, and heat loss due to evaporative cooling were not considered or evaluated.

r U. 'S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 28 8.0      Accident Analyses The fuel consolidation process requires the use of the existing spent fuel pool platform crane for the purpose of ~ moving consolidated fuel' storage.
boxes from the cask laydown area to the spent fuel storage pool. NNECO performed analyses of all postulated scenarios for. the purpose of identifying the consequences of all credible failure modes introduced by this proposed change.
NNECO performed structural analyses to evaluate the results of a fuel assembly or loaded consolidated fuel storage box drop on the spent fuel rack structure. NNECO concluded that the primary function of the rack, which is to maintain separation between the fuel assemblies and insure the flow of the coolant, will not be affected.
Additionally, a structural analysis was performed to evaluate the results of a 6000 pound crane uplif t force on the spent fuel rack structure to simulate a stuck fuel assembly. The resulting stresses were determined to be well within the allowable limits.
A fuel assembly or loaded consolidation box falling on the racks can either fall into a cavity or onto the top of the racks. The resuits of the analyses for each scenario are summarized below.
I l
I

f      <
p
[. .
l    '
l          U 'S. Nuclear Regulatory Commission B12%3/ Attachment 1 Page 29 i
In the case of a fuel assembly falling into an empty cell:
The fuel assembly drop accident was evaluated to determine the effect of the dropped assembly on. the functional and structural integrity of the racks. The analysis indicated that the impact of the fuel assembly on the -
support bars caused plastic deformation of the support bars and 'the fuel l
cell wall supporting the bars. The fuel bundle drop through the rack to the i                fuel support bar resulted in the walls of the rack shearing; however, the    j bundle and support bars did not impact the floor, resulting in no damage to  ,
1 l                the pool liner. (The active fuel length of the bundle will remain contained within the storage rack.)
t In the case of a consolidated !{uel storage box falling into the cavity:
I a
It was conservatively assumed that the impact of the consolidated fuel storage box caused plastic deformation of the support bars and fuel cell wall supporting the bars, resulting in an impact of the consolidated fuel storage box and support bars to the poolliner.
With respect to the racks, the primary function and structural integrity of the racks to maintain separation of the fuel was not impalred.
With respect to the spent fuel pool liner and concrete structural capacity, NNECO completed an evaluation in March 1983 to determine the potential consequences of drops into the Millstone Unit No. 2 spent fuel pool. The I

1 l
U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 30 analysis was performed pursuant to NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." This analysis addressed an object which weighed 67,000 pounds being dropped from a height 41 feet above the spent fuel I                 pool slab through water directly onto the slab at elevation (-)2'-0". The calculation conservatively took no credit for any energy absorption that would occur due to deformation of the object at impact. - Based on this       -l
(
calculation, the dropped 67,000 pound object would be expected - to            I perforate the spent fuel pool liner. The object would penetrate into the concrete floor slab but would not perforate it and therefore would not         j jeopardize the floor slab structural integrity. Also, because no perforation of the concrete slab is expected, no gross leakage from the pool is l
                                                                                                \
expected to occur.
!                Assuming     the   consolidated  storage     box   (weighing   approximately l
2600 pounds), hits the weakest part of the spent fuel pool, which would be the leakage detection system for the liner welds, the leak detection and monitoring system, (as described in the Millstone Unit No. 2 FSAR Sections 5.4.3 and 9.5.2), would be activated. The leak detection and monitoring system would be utilized if (a) a pool liner weld seam was to fail or (b) a local failure of the leak chase collection channel was to occur due to a dropped object directly hitting it. If the spent fuel poolliner is perforated,  '
the five foot thick reinforced concrete spent fuel pool floor slab will act as a water retaining barrier. If either a weld seam or leak chase collection channel was to locally fail due to a dropped consolidated fuel storage box, both overall leakage detection system integrity and pool structural integrity would be maintained with no significant pool water inventory loss.

y j     - U. S. Nuclear Regulatory Commission i      B12563/ Attachment 1 Page 31                                                                                                    l In the case of a fuel assembly falling onto the top of the rack:
l i
l The load resulting from the impact of a dropped fuel bundle onto the top of                          ;
the storage rack was calculated.        Using a finite element model, the                        'l calculated impact force was applied at various locations on the top of the ,
rack.-    The analysis results - in an impact stress that showed local' deformation at the point of contact on the rack. However, the results
                                                                                                                  ]
show that the stresses in the region of active fuel are well below allowable l
stress limits. Therefore, the rack's primary function would remain                               I unaffected.
l l
In the case of a consolidated fuel storage box falling onto the top of the racks:
For the analysis of a loaded consolidated fuel storage box dropping onto the top of the fuel rack structure, a dynamic non-linear multi-spring / mass model of a loaded storage box was developed. This was used to determine the impact load. The maximum impact load resulted in cell wall stresses exceeding the proportional limit. An interactive elastic / plastic analysis of an appropriate region of the rack was therefore performed to evaluate the permanent deformations throughout the region.
The maximum permanent deformation of the cell walls in the active fuel region was found to be 1.5 mils, which is less than 1 percent of the nominal                         ,
clearance between the cell wall and a fuel assembly or consolidated fuel
                                        .-.                                       _         ______-_________:__L

i 4
U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 32 storage box. Hence, this accident would not affect the primary function of the rack which is to maintain separation between fuel assemblies and/or       l consolidated fuel storage boxes and to assure the flow of coolant.
I In the case of a maximum crane uplift force:
An analysis of a typical fuel rack indicated that the force required to 1
deform an individual canister or to overcome the dead weight of the fully '
loaded 3 x 3 temporary storage rack is significantly greater than the load which the spent fuel handling machine can impart. For the analysis, a maximum load of 6000 lbs was assumed. This load was applied to the fuel rack, resulting stresses were well within allowable limits.
I In the case of a consolidated fuel storage box dropped onto an intact fuel assembly in the 3 x 3 rack:
Analysis of this event has determined that the lateral forces transmitted to the rack cell walls due to the bowing of the fuel assembly are sufficient to cause permanent deformation to the cell walls in the storage rack and reduce the separation between the stored assemblies.           However, the   l primary function of the rack (to maintain sufficient separation and to assure coolant flow with a Keff essl   than or equal to .95) is not violated.
The analysis further determined that the drop did not result in any yielding or damage to the stored fuel assembly or adjacent fuel assemblies, resulting in no radiological consequences.

U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 33 l
In' the case of a consolidated it'el storage box drop onto another i
consolidation box or a drop to the cask laydown area floor:
For this analysis it was conservatively assumed that upon impact the consolidated storage box would fail, discharge its content and rupture all of the fuel rods. It was determined that this case would not result in an off-site radiation exposure exceeding the limits of 10 CFR 100 or cause an-Inc ease in the Keff above .95.
In the case of a flow blockage at both ends of a consolidated fuel storage l                                                                                                         i box:
t Analysis determined that the thermal hydraulic aspects of the blocked           l l
consolidated fuel storage box would be. satisfied with only the radial        ']
conduction of heat ~ through the rods and box.
i 9.0   10 CFR 50.59 Summary                                                            ;
l As discussed above, the design of the consolidation equipment meets applicable criteria for normal and abnormal consolidation process operations. The seismic / structural design and analyses of the Consolidated Fuel Storage Box and the 3 x 3 Temporary Fuel Storage Rack meet the l
criteria for normal operating and abnormal conditions for spent fuel in any configuration. Additionally, all the aforementioned postulated accident cases associated with the consolidated fuel storage box and the temporary c:

4 U. S. Nuclear Regulatory Commission B12563/ Attachment 1 Page 34 fuel storage rack will not result in increased site boundary ' doses for Millstone Unit No. 2 fuel handling accidents. Finally, these accidents will not cause an increase in the Keff of the consolidated fuel storage boy or temporary fuel storage rack beyond .95.                                     1 In . addition, based on the above and summarized below, the change associated with the consolidation process does not constitute an
,                unreviewed safety question under 10CFR50.59.
i i                1. The change does not increase the probability of occurrence or the      i consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report because:
l l
                       - The design of the consolidation equipment meets the design criteria for normal and abnormal conditions as delineated in the Millstone Unit No. 2 FSAR Section 9.8.
                       - The seismic structural design of the consolidated fuel storage box l
and 3 x 3 temporary fuel storage rack meet the design criteria for  l l                         normal and abnormal conditions as delineated in the Millstone Unit No.2 FSAR Section 9.8, NRC Standard Review Plan 9.1.2, Regulatory Guide 1.13, and the NRC 1979 Position Paper on Spent 1
Fuel Storage.
                       - The design basis accident reviewed for potential impact due to the l
spent fuel consolidation program is the Fuel Handling Incident (in

                                                                                                      'I
                                                                                                      'i 1
U. S. Nuclear Regulatory Commission B12%3/ Attachment 1 Page 35 j
q the spent fuel pool), FSAR Section 14.19. NNECO concluded that         ;
1 I
the consequences of the assembly drop event are bounded by the FSAR Fuel Handling Incident. The dropping of a consolidated fuel         i storage box is bounded by the cask drop event.        The cask drop event is bounded by a previously analyzed cask drop event. The         1 previous analysis was performed to support the previously approved I
l                               reracking of the Millstone Unit No '2 spent fuel pool.
I
                             - The spent fuel consolidation process equipment does not affect any safety systems that are credited in the design basis.
                             - The MPRT can remove up to 14 fuel rods at once from a fuel l
assembly. If a failure mode associated with the consolidation          I process equipment is assumed, it could be postulated that all l
14 fuel rods fail. The FSAR analysis assumes that 14 fuel rods l                               rupture and that these rods have decayed for 72 hours. Because the fuel rods involved in the consolidation process have decayed for at least five years, the consequences of 14 fuel rod failures during consolidation are bounded by the FSAR analysis.
l l                            - The fuel consolidation process will require some additional fuel i
handling operations. However, the training system, procedures and equipment design have been developed to minimize the likelihood of dropping objects in the consolidation process.         In addition, because of the 5 year decay time, the maximum possible radiological consequences from any fuel handling incident due to

                                          - U. S. Nuclear Regulatory Commission B12%3/ Attachment 1
                                          . Page 36 i
l i
the consolidation process are significantly lower than that l
postulated for the design basis fuel handling accident. For these reasons, it is concluded that there is no impact on the probability of occurrence of the design basis fuel handling accident.
t
                                                                       - The consolidation process will not be performed while plant refueling is in progress. Therefore, the probability of mistoading the reactor core is not increased.                     This will minimize the l                                                                                    probability of consolidating a fuel assembly which was intended to l.
be loaded in the core.
: 2. . The change does not create the possibility for an accident or malfunction of a different type than previously analyzed in the safety analysis report because:
The spent fuel consolidation program and the failure mocies associated with the consolidation process do not modify the plant response to the point where a new accident scenario is created. As                      --
noted earlier, the consolidation process does not affect any safety systems which are credited in the design basis accidents.
Additionally, the offsite does consequences due to a dropped fuel assembly or due to 14 failed fuel rods during the consolidation process are bounded by the FSAR analysis of the fuel handling incident.
_____.m.__.m_._-.__.__.        _,.__m_. _____--..__-.-m_--_m.       --.___-_____m._        _ _ _ _ - _ _ _ - _ m-_ - . -. _ _ _ - . _- m                -   __  -m.-._

y
    .                                                                                             1
                                                                                                'i 4      .                                                                                     I U. S. Nuclear Regulatory Commission B12%3/ Attachment 1                                                                   i Page 37 I
                                                                            ~
                       - The failure moc'es associated .with the program do not represent a new unanalyzed accident. Of the identified failure modes, the two        i i
failures that affect cooling of the consolidated fuel have been         i specifically evaluated to determine ' if. there is potential for creating a new, unanalyzed accident. First, it was determined that l
feel consolidation following a full or partial core offload. is
!                         allowable based on the spent fuel pool heat removal capacity 'or l
l                         capability. Second, a limiting analysis was performed to assess the -
l coolability of a consolidated fuel storage box assuming a complete flow blockage at both ends of the storage box. It was concluded         1 that even with the flow blockage, radial heat conduction is sufficient to remove decay heat and to prevent the fuel and cladding temperature limits from being exceeded.
                       - The cor wildation process does not increase the probability of any accident. The following failure modes were specifically evaluated for an increase in probabiH+v:
Loss of subcritical margin in the spent fuel pool -- NNECO concluded that the technical specification subcriticality rr quire-ment (Keff less than 0.95) is not violated due to these failure m odes.

q e                                                                                               l
            ~
F.            U. S. Nuclear Regulatory Commission B12563/ Attachment 1                                                                   i Page 38 l                            Dropping of heavy loads -- NNECO first determined that use of the       ;
i I
existing spent fuel pool. platform crane to move the five consoll-dated fuel storage boxes during the hot demonstration is accept-l able. NNECO also determined that the major components of the fuel consolidation equipment perform acceptably based on seismic considerations. ' Therefore, the probability of dropping a heavy load in the spent fuel poolis not increased.                                  ;
!                                                                                                   i Misloading the core -- The consolidation process is not performed while plant refueling is in progress. Therefore, the probability of mistoading the reactor core is not increased.
: 3. The change does not reduce the margin of safety as defined in the technical specifications because:
                         - The structural design of the consolic'ated storage box and 3 x 3 temporary storage rack are within the acceptance criteria for safety related components. Additionally, the design of the consoli-dation equipment is. in accordance with all associated industry standards for fuel handling equipment.
                         - As noted above, fuel rod failures during the consolidation process        :
due to a dropped fuel assembly or misoperation of the MPRT are bounded by the assumptions in the FSAR analysis (Fuel Handling Incident). Also, the probability of this incident is not increased due to the consolidation process.

    .                                                                                         \
q U.~ S. Nuclear Regulatory Commission                                                  )
B12563/ Attachment 1 Page 39
                    - It is also postulated that a dropped consolidated fuel storage box could perforate the spent fuel pool liner. This would activate the    !
1 i
spent fuel pool leak detection and monitoring system so that the      I extent of the Jeak would be known. Analysis demonstrates that the 5 foot thick concrete spent fuel pool floor slab would act as a
!                       water retaining barrier. The structural integrity of the pool would I
be maintained with no significant pool water inventory loss.          1 R
t                                                                                              !
l
                    - Analyses were performed to evaluate the coolability of a l
consolidated fuel storage box assuming a flow blockage at the inlet
                                                                                              ]
and outlet. These are the limiting assumptions with regard to consolidated fuel storage box heat removal. These analyses show 1
that the safety limits for fuel and cladding temperature are met.
l l                       Additionally, the more restrictive. criterion for the consolidated spent fuel (i.e., cladding temperatures less than or equal to 6500F) are also met.                                                          I
                    - The technical specification requirement of Keff ess   l  than 0.95 is not violated due to any failure mode related to the spent fuel consolidation process.
j       Based on the above, NNECO believes that the consolidation process is safe and I       technically acceptable and that the change associated with the consolidation l       process does not constitute an unreviewed safety question under the provisions of 1
i 10CFR50.59.
l t
j l
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