ML20151M903
| ML20151M903 | |
| Person / Time | |
|---|---|
| Site: | 07003055 |
| Issue date: | 07/31/1988 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20151M897 | List: |
| References | |
| NUDOCS 8808080032 | |
| Download: ML20151M903 (31) | |
Text
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1 4
VOGTLE ELECTRIC GENERATING PLANT - UNIT 2 SPECIAL NUCLEAR MATERIAL O
-l LICENSE APPLICATION REVISION 1 REVISION INSERTION INSTRUCTIONS JULY 1988 Page Action Title Page Replace Distribution List Replace i
Replace 111 Replace 1-3 Replace 1-4a Add 1-5 Replace 1-11 Replace 1-19 Replace 1-20 Replace 2-7 Replace 2-9 Replace 2-10a Add 2-13 Replace
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2-15 Replace 2-20 Replace 2-21 Replace 2-22 Replace F-1 Replace Effective Page List Tab Add Effective Page List Add 0
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APPLICATION FOR SPECIAL NUCLEAR MATERIAL LICENSE
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.i FOR j
RECEIPT, POSSESSION AND STORAGE OF UNIRRADIATED REACTOR FUEL AND ASSOCIATED RADIOACTIVE MATERIALS i
'l FOR i
VOGTLE ELECTRIC GENERATING PLANT 1
UNIT 2 i
Revision 1 l
O July 1988 1
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- 0 GEORGIA POWER COMPANY l
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VOGTLE ELECTRIC GENERATING PLANT-UNIT'2 SPECIAL NUCLEAR MATERIAL LICENSE APPLICATION Distribution List Manual Holder Copy Number R.
J.
Florian 1
B.
E.
Hunt 2
R.
A.
Thomas 3
J.
E.
Jeiner, Erq.
4 B.
W.
Churchill, Esq.
5 R.
T.
Oedamer 8
J.
A.
Bailey 9
B.
R. Quick 10
~
D.
R.
Marnon 11 I
G.
Bockhold, Jr.
12 W.
E.
Burns 13 r
s i
r I
(
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,O C626n Revision 1 7/88 i
3
3..;
2t TABLE OF CONTENTS 6
Page 1.0 General Information 1-1 1.1 Reactor and Fuel 1-1 1.i.1 General 1-1 1.1.2 Fuel Assemblies 1-1 1.1..)
Aseembly Enrichmento and Weights 1-1 1.?.4 Total Fuel Assemblies and Uranium 1-2 1 'd Stctsge Conditions 1-2
- 1. 2. '.
Fuel Storage Area 1-2 1.2.2 Euel Storage Area Environment 1-2 1.3.3 Fuel Storage Area Activities 1-2 7.2.4 Juel Handling and Auxiliary i
nuilding Equipment and Systems 1-3 1.1.I Fire Alarm and Control Systems 1-7 1.2,6 A.ccess Control 1-8 1.3
.9hysical Protection 1-8 r~T 1.4 Trinefer of Special Nuclear Material (SNM) l-G b
1.4.1 Shipment to VEGP 1-8 1.?.2 Shipment from VEGP 1-9 1,. 3 CNM Control and Accounting 1-9 1.5 Financial Protection and Indemnity 1-10 1.5.1 Proof of Financial Protection 1-10 1.S.2 Not Applicable.
1-10 1.5.3 Not Applicable.
1-10 2.0 Health and Safety 2.1 Radiation Control 2-1 2.1.1 Radiation Safety Personnel Minimum 2-1 Qualifications 2.1.2 Radiation Safety Personnel 2-1 Responsibilities 2.1.3 Radiation Safety Personnel Training 2-2 and Experience 2.1.4 Radiation Contamination Equipment and 2-2 Procedures 2.1.5 Calibration and Testing of 2-3 Radiological Protection Inetruments
()
2.1.6 Compliance with 10 CFR 20 2-4 2.1.7 Disposal of Radioactive Wastea 2-4 i
TABLE OF CONTENTS (Continund)
~,
Page g
2.2 Nuclear Criticality Safety 2-5 2.2.1 Nuclear Criticality Safety and Fuel 2-5 Handling Personnel Qualifications 2.2.2 Nuclear Criticality Safety and Fuel 2-5 Handling Personnel Responsibilities 2.2.3 Shipping Container 2-6 2.2.4 Fuel Assembly Storage 2-6 2.2.5 Criticality Safety-Enrichment 2-10 2.2.6 Criticality Safety-Neutron Absorbers 2-10 2.2.7 Criticality Safety-Moderation Control 2-11 2.2.8 Criticality Safety-Methodology 2-12 2.2.9 Criticality Safety-Fuel Movement 2-12 and Inspection 2.2.10 Exemption from 10 CFR 70.24 2-13 2.3 Accident Analysis 2-13 3.0 Other Material Requiring NRC License 3-1 3.1 Description of Material and Conditions 3-1 of Storage 3.1.1 Type and Amount of Material 3-1 3.1.2 Conditions of Storage 3-1 3.2 Use Other Than Storage 3-1 3.3 Radiation Protection 3-1 3.4 Control and Accountability 3-2 Appendix A Initial Fuel Receipt Fire Protection A-1 Area Boundary Description and VEGP FSAR Subsection 9A.1.2, Fire Area 1-A3-LD-B Appendix B VEGP FSAR Subsection 13.1.3 and Key Health B-1 Physics and Criticality Personnel Resumes Appendix C Nuclear Energy Liability Insurance C-1 Association (ANI) Liability Policy No. NF-302; Mutual Atomic Energy Liability Underwriters Liability Policy No. MF-129 Appendix D VEGP FSAR Paragraph 13.2.2.1.1 D-1 Appendix E Westinghouse Electric Corporation E-1 Criticality Analysis Methodology Appendix F Holtec International Criticality Analysis lh Methodology F-1 Effective Page List l
11 Revision 1 7/88
\\
4 LIST OF TABLES Table Page 1.1-1 Weight of U-235 in Fuel Assemblies 1-11 1.2-1 Spent Fuel Rack Storage Cell and Module Data 1-12 2.2-1 Spent Fuel Rack Design Data 2-15 8
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iii
LIST OF FIGURES O
Figure Page 1.1-1 VEGP FSAR Figure 4.2-1, 17 x 17 1-13 Standard Fuel Assembly Cross Section 1.1-2 VEGP FSAR Figure 4.2-2, 17 x 17 Standard 1-14 Fuel Assembly Outline 1.1-3 VEGP FSAR Figure 4.2.3, Fuel Rod Schematic 1-15 1.2-1 Fuel Handling Building Layout 1-16 1.2-2 Unit 2, New and Spent Fuel Storage Area 1-17 Layout 1.2-3 VEGP ESAR Figure 9.1. 5-1, Spent Fuel Cask 1-18 Bridge Crane Clearance and Hook Travel Envelope (Plan View) 1.2-4 Unit 2 Spent Fuel Pool Storage Rack 1-19 Typical Layout for Storage of New Fuel 1.2-5 Unit 2 Spent Fuel Racks New Fuel Storage (Dry) 1-20 2.2-1 VEGP FSAR Figure 9.1.1-1, New Fuel Rack 2-16 2.2-2 New Fuel Storage Rack Installation and 2-17 Assembly 2.2-3 New Fuel Storage Rack Cell Geometry, 2-18 Top View 2.2-4 Typical Spent Fuel Rack 9 x 9 Module 2-19 2.2-5 Typical Spent Fuel Rack Storage Cell 2-20 Interconnection, Top View 2.2-6 Unit 2 Spent Fuel Rack Storage Cell Geometry, 2-21 Top View 2.2-7 Typical Spent Fuel Rack Storage Cell 2-22 Elevation, Side View O
iv Revision 1 7/88
1.2.4 FUEL HANDLING AND AUXILIARY BUILDING EQUIPMENT AND SYSTEMS The fuel handling building and auxiliary building are Seismic Category 1, reinforced concrete structures common to the two-unit plant.
The fuel handling building consists of two spent fuel storage pools, a new fuel storage pit, a cask loading pit, fuel transfer canals, and a cask washdown area.
The new fuel receipt, unloading, and shipping container storage area of the auxiliary building has railroad access at grade, and crane facilities for receiving new fuel.
The crane system is capabia of transferring the new fuel into the fuel handling building.
Figure 1.2-1, page 1-16, shows the general layout of the fuel handling building and the new fuel receipt, unloading, and shipping container storage area of the auxiliary building.
The new fuel storage pit is a separate reinforced concrete pit providing temporary dry storage for new fuel assemblies, thus satisfying the new fuel storage requirements for both Units 1 and 2.
The storage pit is
['T protected from the effects of natural phenomena, k/
including earthquakes, winds, tornadoes, floods, and external missiles, by the Seismic Category 1 fuel i
handling building.
The new fuel storage racks are designed to include storage for 162 new fuel assemblies placed in a vertical upright position with a minimum center-to-center spacing of 21 in.
This geometry ensures that kefg shall not exceed 0.95 for fresh fuel assemblies with up to 3.5 w/o U-235 enrichment even at optimum moderation conditions.
All surfaces that come into contact with the fuel assemblies are made of annealed austenitic stainless steel; whereas, the supporting structure may be painted 4
carbon steel.
j i
The new fuel storage racks are designed to withstand nominal operating loads as well as safe shutdown earthquake (SSE) and operating basis earthquake (OBE) seismic loads.
The new fuel storage racks are designed to meet the Seismic Category 1 requirements of Regulatory Guide 1.29.
The new fuel storage racks are also designed to withstand the maximum uplift force of the fuel handling g
machine (4000 lb)-
The new fuel storage pit is covered p/
(_
to prevent objects from falling on the racks. The cover is designed so that it will not fall during a seismic l-3
event and damage the fuel racks.
Administrative controls are used when a section of the protective cover is opened for handling of the new fuel assemblies.
The spent fuel storsge facility is designed to meet the guidelines of ANS 57.2 and is' located within the Seismic Category 1 fuel handling building.
The facility is protected from the effects of natural phenomena, such as earthquakes, winds, tornadoes, floods, and external missiles.
The f acility $ s designed to maintain its structural integrity following an SSE and to perform its intended function following a postulated hazard, such as fire, internal missiles, or pipe break.
Each unit is provided with its own spent fuel pool.
Each spent fuel pool is approximately 41-ft deep, constructed of reinforced concrete, and lined with 1/4-in.-thick etainless steel.
The spent fuel storage racks to be installed in the Unit 2 spent fuel pool prior to new fuel receipt consist of at least two free standing modules as shown jn figure 1.2-4, page 1-19.
The number of spent fuel racks and their arrangement in the spent fuel pool may be different from that chown in figure 1.2-4, depending on the delivery schedulo of spent fuel racks.
For dry storage conditions, new fuel storage in the Unit 2 spent fuel racke will be restricted to two 11 by 10 rack modules (B-1 and B-2) which will be positioned such that they are separated from each other, the pool walls, and any other objects in the pool by at least 12 in.
Table 1.2-1, page 1-12, provides the cell count and module I.D.
data.
The storage cells consist of 8.75 inch nominal prismatic openings fermed by seam welding precision formed channels.
Thase cells are interconnected using longitudinal angle connectors to form a honeycomb construction structure.
Each fuel storage location incorporates the Boraflex neutron absorbing material (boron carbide powder uniformly dispersed in a polymeric matrix) which is held in place by a stainless steel sheathing.
The sheathing is held in place by a combination of spot welds and an auxiliary retainer as shown in figure 2.2-6, page 2-21.
The auxiliary retainer is constructed from a sheet of type 304 stainless steel that is 8 in, wide, 20 gauge thick, a"d 165 in. long by forming it into the shape shown in figure 2.2-6.
The preserce of the auxiliary retainer was considered in the criticality safety analysis.
The Boraflex encasement method provides for unconstrained in-plane contraction (or expansion) of the poison material and lands complete lateral support to it to protect it from slumping.
The material is not sealed since it is compatible with the pool environment, g
1-4 Revision 1 7/88
The spent fuel storage cell walls, as well as all other structural components, are fabricated from Type 304L
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(_)
stainless steel.
The nominal cell pitch is 10.58 in. in l
the north-south direction, and 10.40 in. in the east-west direction.
r These free standing, self-supporting modules are designed to ensure that, for wet storage conditions, the neutron multiplication factor (keff) shall not exceed 0.95 with fresh fuel assemblies of up to 4.55 w/o U-235 enri:hment in all storage locations, even in the event the pool is flooded with unborated water.
The design also ensures that, for dry storage conditions, k gg e
shall not exceed 0.95 with fresh fuel assemblies of up to 4.55 w/o U-235 enrichment in all storage locations even at optimum moderation conditions.
1 1-4a Revision 1 7/88
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1-4b Revision 1 7/88
The design and safety evaluation of the spent fuel racks is in accordance with the NRC position paper, "Review rk' and Acceptance of Spent Fuel Storage and Handling Applications."
The racks, which are Nuclear Safety Class 3 and Seismic Category 1 structures, are designed to withstand normal and postulated dead louds, live loads, loads resulting from thermal effects, and loads caused by OBEs and SSE events.
For dry storage conditions, new fuel storage in the Unit 2 spent fuel racks will be restricted to two 11 by 10 rack modules (B-1 and B-2) where minimum l
separation distances between the rack modules and the pool walls are sufficient to ensure no impact with each other, other rack modules or the pool walls in the event of an OBE or SSE event.
Fuel handling procedures shall provide the necessary guidance and control to ensure that, for dry storage conditions, new fuel assemblies stored in each of these two rack modules are placed only in storage locations designated as acceptable for seismic events, as shown in Figure 1.2-5 page 1-20.
The racks are also designed with adequate energy absorption capabilities to withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel handling machine.
Handling equipment (spent f-s fuel cask bridge crane) capable of carrying loads
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heavier than a fuel assembly are physically prevented by design from traveling over the spent fuel storage area.
The fuel storage racks can withstand an uplift force' (4000 lb) equal to the uplift capability of the fuel handling machine.
All materials used in construction are compatible with the storage pool environment, and all surfaces that come into contact with the fuel assemblies are made of annealed austenitic stainless steel.
All the materials are corrosion resistant and will not contaminate the fuel assemblies or pool environment.
Design of the facility, which is in accordance with Regulatory Guide 1.13, ensures adequate safety under normal and postulated accident conditions.
The following fuel handling equipment and components may be used to handle, inspect, and place fuel into storage until initial core loading:
1.
Scent Fuel Cask Bridge Crane The spent fuel cask bridge crane is used to transfer new fuel assemblies between the shipping / receiving
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area of the auxiliary building and the new fuel 1-5 Revision 1 7/88
storage pit of the fuel handling building.
By means lll of built-in limit switches, the crane is restricted from traveling near or over the new fuel storage pit when the main hoist is handling loads in excess of 15 tons, as shown in figure 1.2-3, page 1-18.
This crane has a 125-ton capacity main hoist which is designed to.be single-failure-proof in accordance with NUREG-0554.
A 15-ton auxiliary hoist is provided on the same trolley, and a 2-ton monorail hoist is provided on the bridge.
The auxiliary hoist and the monorail hoist are used in new fuel handling and maintenance tasks.
The path of the crane does not pass over either of the spent fuel storage pools.
Positive restraints are provided on the crane to prevent the bridge, trolley, or any other part from becoming dislodged and falling on structures or equipment situated below the crane in the event of an SSE.
The crane is capable af retaining the maximum design load during an SSE, although the crane is not qualified to operate following the SSE.
The spent fuel cask bridge crane is an all-steel construction, electric overhead, top running, double box girder, motorized dual drive bridge crane.
The crane is mounted on two parallel runway rails traversing between the auxiliary and fuel handling buildings.
g 2.
Fuel Handling Machine The fuel handling machine, which is a wheel-mounted walkway spanning the spent fuel pools, carries two trolley-mounted electric hoists (load rated at 4000 lb), each located on an overhead st'ructure.
This machine is used for handling fuel assemblies within the fuel storage area by means of a handling tool suspended from the hoist.
The fuel handling machine is equipped with a load monitor for monitoring fuel assembly loads.
The bridge and hoist speeds are variable.
The approximate maximum speed for the bridge is 33 ft/ min and for the hoist is 20 ft/ min.
Hoist limit switches and interlocks prevent loads in excess of 2300 lb from being lifted.
3.
New Fuel Elevator The new fuel elevator (load rated at 2000 lb) consists of a box-shaped elevator assembly with its top end open and sized to ho~use one fuel assembly.
The elevator is used exclusively to lower a new fuel assembly to the bottom of the spent fuel storage area where the assembly is transported to the 1-6
TABLE 1.1-1 f-WEIGHT OF U-235 IN FUEL ASSEMBLIES t
U-235 Contained (a)
Description (kg) 65 Region 1 @ 2.1 w/o 648 64 Region 2 @ 2.6 w/o 786 64 Region 3 @ 3.1 w/o 935 193 = TOTAL ASSEMBLIES 2369 = TOTAL kg i
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The above quantities of U-235 were calculated using a nominal fuel assembly uranium weight of 463.4 kg with the maximum manufacturing enrichment tolerance of.05 w/o U-235 applied to each region.
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SPENT FUEL RACK STORAGE CELL AND MODULE DATA NUMBER OF CELLS Total Module No. of Number of East-West North-South No.of I.D.
Modules Cells per Module Direction Direction Cells A
4 99 11 9
396 B
3 110 11 10 330 O
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1-12 Revision 1 7/88
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SNM LICENSE APPLICATION FIGURE 1.2-5 10 9
8 7
X X
X X
X 6
X X
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X RACK MODULE 5
X X
X X
X B-1 (11x10) 4 X
X X
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RACK MODULE 5
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2 1
A B C D E F G H J K L "X" LCCATIbro J.RE PERMITTED TO STORE NEW FUEL ASSEMBLIES IN A DRY POOL, "BLANK" LOCATIONS MUST NOT STORE FUEL ASSEMBLIES IN A DRY POOL.
h Revision 1 7/88 UNIT 2 SPENT FUEL RACKS vocTLE GeorgiaPower A SZ"$o["dr"$UNNNT NEW W EL STORAGE ( DR'l) d33 9 1-W6L
l typical.
The center-to-center spacing between individual storage cells in each cell group is 21.00 in.
((s) typical.
3 Cue center-to-center spacing between the closest of adjacent group cells is 51.75 in, typical.
New fuel storage cell wall thickness is 0.075 in, typical.
New fuel will be held off the floor of the new fuel vault by resting on support plates at the bottom of each cell.-
The east side of the racks measure nominally 17.5 in, from the east vault wall to the center of the fuel cells.
The west side of the racks measures nominally 17.5 in, frem the west vault wall to the center of the fuel cells.
There is at least 12 in. to the center of the closest fuel cell as measured from the south end of the racks to the south vault wall.
There is at least 12 i
in, to the center of the closest fuel cell as measured l
from the north'end of the racks to the north vault wall.
No neutron poisons are used in the fabrication of the new fuel storage racks.
The spent fuel storage racks are fabricated in the form of eight modules in three discrete sizes.
A typical rack module is shown in figure 2.2-4, page 2-19.
As can be seen from the pictorial illustration in that figure, the storage cells are interconnected to form a honeycomb structure.
The cell interconnecting angles can be I~/
clearly seen in figure 2.2-5, page 2-20, which shows a 4 h
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by 4 cell array in a cross sectional view.
Figure 2.2-6, page 2-21, shows the poison emplacement scheme along with the cell interconnecting angles in a 2 by 2-array.
The elevation of a typical storage location can be seen in figure 2.2-7, page 2-22.
The nominal spacing between two adjacent modules is 1.5 in., and the minimum peripheral rack-to-wall gap is 2-1/8 in.
For dry storage conditions, new fuel' storage in the Unit 2 spent fuel racks will be restricted to two 11 by 10 rack l
modules (B-1 and B-2) where minimum separation distances l
between the rack modules and the pool walls are i
sufficient to ensure no impact with each other, other rack modules, or the pool walls in the event of an OBE or SSE event.
Fuel handling procedures shall provide the necessary guidance and control to ensure that, for dry storage conditions, new fuel assemblies stored in
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each of these two rack modules are placed only in j
storage locations designated as acceptable for seismic events, as shown in Figure 1.2-5 page 1-20.
The details of the module layout may be found in Figure 1.2-4, pages 1-19.
Table 2.2-1, page, 2-15, gives the essential spent fuel rack design data.
The integrity of new and spent fuel storage racks is discussed in subsection 1.2.4 of this application.
2-7 Revision 1 7/88
s Administrative and fuel handling procedures shall provide the necessary guidance and control for movement of assemblies from their respective receipt container to a storage rack location.
An outline of the fuel handling procedure is as follows:
1.
Fuel assemblies will be received on site in Westinghouse model RCC-1 or RCC-3 shipping containers presently licensed under NRC License No. USA /5450/AF.
2.
The vehicle and all packages will be surveyed by Health Physics Department personnel in accordance with the requirements of 10 CFR 20.205 and approved Health Physics procedures for the receipt of new fuel.
3.
The transport truck and metal shipping containers will be moved to the unloading area of the auxiliary building.
4.
The metal shipping containers will usually be unloaded from the transport truck using the spent fuel cask bridge crane, moved to the floor of the new fuel unloading area, and placed in a suitable unloading attitude.
The shipping containers may be lll unloaded from the new transport truck using a suitable mobile crane or fork lift truck and transferred to the new fuel receipt area of the auxiliary building.
5.
Before opening a shipping container, an external inspection for damage and a pressure differential check will be made.
6.
Health Physics Department personnel will ensure that proper radiological controls are established for the opening of the shipping container in accordance with the requirements of 10 CFR 20.205 and Health Physico procedures for the receipt of new fuel.
7.
The cover will be removed from the metal shipping container and set aside.
8.
An inspection of the inside of the container will be made and documented.
9.
Shipping container outrigger members (if so equipped) located on the outside of the container assen61y will be positioned as required for a
unloading.
Other necessary support structures will W
be positioned, as required, for pivoting the 2-8
d gs support frame of the shipping container to the
(_)
vertical position.
Fuel assemb2y clamping frames, except top and bottom, may be partially loosened at this time.
10.
The support frame will then be slowly raised to the vertical position where the supports and securing hardware will be installed to secure the support frame in the vertical position.
11.
The new fuel assembly handling tool and a scale for the spent fuel cask bridge crane will be installed and the new fuel assembly handling tool will be latched to the fuel assembly upper nozzle.
f 12.
After lifting cable slack is removed, fuel assembly j
shipping container clamping frames can be removed.
l The fuel assembly may now be removed from the shipping container.
I 13.
The fuel assembly protective cover will be inspected for damage and removed.
An inspection o
j and radiation survey will be performed on the l
exposed assembly.
Protectiva covers may then be replaced.
To assure drainage of water that may accumulate around the stored fuel assemblies, the O
protective covers will remain open on the bottom or a slit in.the protective cover will be placed near the b6ttom of the fuel assembly.
14.
The new fuel will be placed in storage by one of the following methods:
a.
The assembly will be inserted into the new fuel storage racks using the new fuel handling tool and the spent fuel cask bridge crane.
The fuel handling machine, using the new fuel handling tool, will remove the assembly from the new fuel storage rack location and transport it to the new fuel elevator.
The fuel assembly will then be placed in the new fuel elevator and lowered into the Unit 2 spent fuel storage pool.
The spent fuel handling tool and the fuel handling machine will be used te place the new fuel assembly into an acceptable Unit 2 spent fuel storage rack location.
The rack position of each fuel assembly will be recorded i
as it is placed; or l
I b.
The assembly will be inserted into the new fu.el I
storage racks using the new fuel handling tool 3
and the spent fuel caak bridge crane.
The rack y_)
position of each fuel assembly will be recorded as it is placed.
2-9
15.
Fuel assembly components, such as rod control cluster assemblies, burnable poison rod assemblies, thimble plugs, or source assemblies, may be visually inspected at this time.
16.
Steps 11 through 15 above will be repeated for unloading the second fuel assembly from the shipping container.
After the fuel assemblies have been removed from the shipping container, the container will be made ready for return shipment to Westinghouse.
17.
The above procedure will be repeated until all the fuel has been unloaded, inspected, and properly stored.
2.2.5 CRITICALITY SAFETY-ENRICHMENT The nuclear criticality safety of the VEGP new fuel assembly storage is based upon values that bound the maximum U-235 enrichments.
The maximum U-235 enrichment which would be expected due to manufacturing tolerances is 3.15 w/o.
The Unit 2 spent fuel storage racks are designed for wet or dry storage of fresh 17 x 17 Westinghouse standard fuel assemblies.
Each fuel storage cell incorporates a neutron-absorbing material (boron carbide powder uniformly dispersed in a polymeric matrix) which, in conjunction with llh the nominal center-to-center cell spacing of 10.58 in. in the north-south direction, and 10.40 in. in the east-west di rec t'i on, ensures that, for wet storage conditions, keff shall not exceed 0.95 with fresh fuel assemblies of up to 4.55 w/o U-235 enrichment in all storage cell locations, even in the event the pool is flooded with unborated water.
The kef f calculation was based on deviations from nominal, due to manufacturing, that provide a conservative value of the calculated keff.
The design also ensures that, for dry storage conditions, keff shall not exceed 0.95 with fresh fuel. assemblies of up tc 4.55 w/o U-235 enrichment in all storage cell locations even at optimum moderation conditions.
The new fuel storage racks are designed for dry storage of new fuel assemblies at a minimum center-to-center spacing of 21 in.
This spacing is sufficient to ensure that keff shall not exceed 0.95 for fresh fuel assemblies of up to 3.5 w/o U-235 enrichment even at optimum moderation conditions.
O 2-10 Revision 1 7/88
2.2.6 CRITICALITY SAFETY-NEUTRON' ABSORBERS The neutron-absorbing material, Boraflex, used in the spent fuel rack construction is manufactured by Brand Industrial Services, Inc. and is fabricated to the safety-related nuclear criteria of 10 CFR 50, e
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2-10b Revision 1 7/88
l 2.2.10 EXEMPTION FROM 10 CFR 70.24 O
Pursuant to 10 CFR 70.24(d), it is requested that Georgia Power Company be exempted from the monitoring requirements of 10 CFR 70.24(a) as they apply to the handling and ctorage of nuclear fuel assemblies at VEGP.
The procedures and storage facilities described in this application provide assurance that inadvertent criticality cannot occur during receipt, handling, and storage of nuclear fuel assemblies.
The Westinghouse shipping containers, models RCC-1 and RCC-3, are approved for transport of Fissile Class II materials as defined in 10 CFR 71.4.
The procedures for unloading and inspecting the fuel are based in part upon the prevention of criticality during these operations.
The new fuel storage racks are designed to prevent keff from exceeding O.95 for fresh fuel assemblies of up to 3.5 w/o U-235 enrichment,even at optimum moderation conditions.
The spent fuel storage racks are designed to prevent keff from exceeding 0.95 for fresh fuel assemblies of up to 4.55 w/o U-235 enrichment, even at optimum moderation conditions.
New fuel stored at VEGP will be 3.15 w/o U-235 enrichment or less (including manufacturing tolerances), and only one fuel assembly will be out of a shipping container or storage rack at any given time (3
\\_/
It is also requested that Georgia Power Company be exempted from the requirements of paragraph 70.24(b) in accordance with paragraph 70.24(c), since VEGP Unit 1 is operating under Operating License NPF-68 and Unit 2 is being built under Construction Permit No. CPPR-109.
2.3 ACCIDENT ANALYSIS The new fuel storage racks were designed to preclude storage of a fuel assembly other than where intended, to withstand the uplift force which could occur due to a fuel assembly hanging up during lifting, and to withstand the impact load of a dropped fuel assembly.
Each fuel assambly stored in the new fuel storage rack is protected from falling objects by steel cover plates, which are administratively controlled when removed and are designed not to fall and damage the fuel racks during a seismic event.
The spent fuel storage racks were designed with' adequate
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energy absorbtion capabilities to withstand the impact j
of a dropped fuel assembly from the maximum lift of the fuel handling machine.
Fuel assemblies will be further g
protected from falling objects, because the only other
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fuel building crane (spent fuel cask crane) capable of carrying loads heavier than a fuel assembly is prevented i
2-13 l
l
from physically traveling over the spent fuel storage area.
The spent fuel storage racks can withstand an lll uplift force (4000 lb) equal to the uplift capability of the fuel handling machine.
Because of the close spacing of the cells, it is impossible to insert a fuel assembly in other than design locations.
Inadvertent insertion of a fuel assembly between the rack periphery and the pool wall or placement of a fuel assembly across the top of a fuel rack is considered a postulated accident, and as such, realistic initial conditions can be taken into account.
The new fuel and spent fuel storage racks are designed to withstand normal and postulated dead loads, live loads, loads resulting from thermal effects, and loads caused by the operating basis earthquakes and safe shutdown earthquake events.
For dry storage conditions, new fuel storage in the Unit 2 spent fuel racks will be restricted to two 11 by 10 rack modules (B-1 and B-2) l where minimum separation distances between the rack modulos and the pool walls are sufficient to ensure no impact with each other, other rack modules or the pool walls in the event of an OBE or SSE event.
Fuel handling procedures shall provide the necessary guidance and control to ensure that, for dry storage conditions, new fuel assemblies stored in each of these two rack modules are placed only in storage locations designated as acceptable for seismic events, as shown in Figure 1.2-5 page 1-20.
In the event of an accident, the plan of ac. tion would be to follow the Vogtle Electric Generating Plant Emergency Plan and to ta,ke specific recommended actions depending upon the severity of the accident.
The Health Physics Department is responsible for checking for any resulting contamination and for taking appropriate decontamination steps, if required.
2-14 Revision 1 7/88
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SPENT FUEL RACK DSSIGN DATAk Fuel Type:
Westinghouse 17 x 17 Rack Type:
Honeycomb construction; poisoned high density Nominal Center-to-Center Spacing:
l North-South:
10.58 East-West:
10.40 Cell inside dimension:
8.75 Cell wall thickness:
0.075 Poison length:
139 i
Poison width:
7.75 Poison thickness:
0.075 4
Minimum B ' Loading:
0.020 ( gm/cm )
1 2
Poison sheathing
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0.02 l
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All dimensions in inches unless otherwise indicated.
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CRITICALITY ANALYSIS METHODOLOGY The high density spent fuel storage racks for the Vogtle Electric Generating Plant Unit 2 are designed to assure that the neutron multiplication factor (keff) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at a temperature corresponding to the highest reactivity.
The maximum calculated reactivity includes a calculational bias and a margin for uncertainty in reactivity calculations and in mechanical tolerances.
The uncertainties in reactivity calculations and mechanical tolerances are statistically combined such that the true k egf will be equal to or less than 0.95 with a 95 percent probability at a 95 percent confidence level.
Applicable codes, standards, and regulations, or pertinent sections thereof, include the following:
General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.
USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, New Fuel Storage, and Section 9.1.2, Spent Fuel Florage.
O USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.
USNRC Regulatory Guide 3.41, Validation of Calculational Methods for. Nuclear Criticality Safety (and related ANSI N16.9-1975).
ANSI /ANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
To assure the true reactivity will always be lees than the calculated reactivity, the following conservative assumptions were made:
Moderator is pure, unborated water at a temperature within the design basis range corresponding to the highest reactivity.
F-1
Lattice of storage racks is assumed infinite in all directions; i.e.,
no credit is taken for axial or radial neutron leakage (except in the assessment of certain abnormal / accident conditions).
Neutron absorption in minor structural members i's neglected; i.e.,
spacer grida are replaced by water.
The fuel of highest anticipated reacti~vity is assumed; i.e.,
fresh unburned fuel of 4.55 w/o enrichment, in the Westinghouse 17 X 17 standard fuel assembly geometry.
l Criticality analyses of the high density spent fuel storage racks were performed with the AMPX-KENO computer package, using 1
the 27-group SCALE cross-section library with the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treat:aent).
Benchmark calculations indicate a bias of 0.0106 + 0.0048 (95%/95%).
In the geometric model used in KENO, each fuel rod and its cladding were described explicitly and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the centerline of the water-gap between storage cells.
These boundary conditions have the effect of creating an infinite array of storage cells in all directions.
The CASMO-2E computer code, a two-dimensional multigroup transport theory code for fuel assemblius, has also been benchmarked and was used both for verification calculations and as a means of evaluating small reactivity increments associated with manufacturing tolerances.
CASMO-2E benchmarking resulted in a calculational bias of 0.0013 + Q.0018 (95%/95%).
- However, limitations in the geometry options available in CASMO-2E required minor approximations in the geometric description (e.g.,
in the description of the aoraflex absorber and the use of an average water-gap thickness) which apparently contributes to a small over-prediction in the absolute value of the CASMO cell infinite muliplication factor.
l A third independent method of critically analysis, utilizing diffusion / blackness theory, was also used for additional confidence in results of the primary calculational methods, although no reliance for criticality safety is placed on the reactivity value from the diffusion / blackness theory tect igue.
Thic technique, however, is used for auxiliary calculaticss of the cmall incremental reactivity effect of eccentric fuel positioning that would otherwise be lost in normal KENO statistical variations, or would be inco,sistent with CASMO-2E 1.
SC?; '- is an acronym for Standardized Computer Analysis for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.
F-2 Revision 1 7/88
f.
VOGTLE ELECTRIC GENL7ATING PLANT - UNIT 2 SPECIAL NUCLEAR MATERIAL
,_s LICENSE APPLICATION s_-)
(,
EFFECTIVE PAGE LIST l
All text pages, tables, and figures of the Vogtle Electric Generating Plant - Unit 2 Special Nuclear Material License Application should be considered Revision O except for the following which are Revision 1.
Title page Distribution list ii v
1-4 1-4a 1-4b 1-5 1-12 (Table 1.2-1) 1-19 ( Figure 1. 2-4) 1-20 (Figure 1.2-5) 2-7 2-10 2-10a
(~'T 2-10b l_/
2-14 2-15 (Table 2.2-1) 2-20 ( Figure 2. 2-5 )
2-21 (Figure 2.2-6) 2-22 (Figure 2.2-7)
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