ML20235G201
| ML20235G201 | |
| Person / Time | |
|---|---|
| Site: | 07003013 |
| Issue date: | 05/06/1987 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20235F489 | List: |
| References | |
| 0327A, 327A, NUDOCS 8707140151 | |
| Download: ML20235G201 (35) | |
Text
.
o TABLE OF CONTENTS.
Chapter Page 1
General Information.
1 1.1 Reactor and Fuel..............
1 1.2 Storage Conditions.
11 1,3 Physical Protection.
19' 1.4 Transfer of Special Nuclear Material.
19 4
1.5 Financial Protection and Indemnity..'.....
22 2
Health and Safety..
22-2.1 Radiation Control...............
22 2.2 Nuclear Criticality. Safety..........
24' 2.3 Accident Analysis.
30 4
-\\
}
3 Other Materials Requiring NRC License.
31
]
1 ATTACHMENTS l
Attachment A Regulatory Guide 3.15, Revision 1, April 1983 Attachment B Byron /Braidwood Fire Protection Report, Sections 2.3.12, 2.3.12.1, 3.2(1), 3.7(1)
{
l Attachment C Commonwesith Edison Company Nuclear Procedures Nos. I tneough 8
]
Attachment D Training and experience of personnel responsible for radiation safety.
Attachment E Commonwealth Edison Company Radiation Protection Standards Attachment F Sections 9.1.1 and 9.1.2 of NUREG-0876, Safety Evaluation Report related to the operation of Byron Station, i
Units 1 add 2 Attachment G Byron /Braidwood FSAR, Sections 9.1.1.1, 9.1.1.3, 9.1.2.1,
-l 9.1.2.3
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LIST OF FIGURES AND TABLES Figure
-Page 1.1-1 B/B;FSAR Figure 4.2-1,17x17 Optimized Ibel i
Assembly Cross Section.
2
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1.1-2 B/B FSAR Figure 4.2-2,.17x17 Optimized'Tuel 1
Assembly Outline..
3 i
1.1-3 B/B FSAR Figure'4.2-3, Fuel Rod Schematic.
4 1.1-4 B/B.FSAR Figure 4.2-6, Top Grid to Guide Thimble'and-
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Top Nozzle Attachment.
7' y
- 1. 1-5 B/B FSAR. Figure 4.2-7, Guide Thimble to 33ottom Grid i
and Nozzle Joint.
8
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1.2-1 M-27 Sheet 7, Fuel Handling Building.
12 1.2-2 B/B FSAR Figure 9.1-1, New Fuel Rack Installation.
- 13.
1.2-3 B/B FSAR Figure 9.1-2, Spent Fuel Storagt: Rack Arrangement.
14 J
1.2-4 B/B FSAR Figure 9.1-3, Typical Spent Fuel Rack Isometric.
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I Table Page 1.1.2 General Fuel Data 5
i 1.4.3 Indei of Nuclear Procedures 21 q
i iv 0661m(011787) i 0327A
Chapter 1 General Information 1.1 Reactor and Fuel 1.1.1 Authorization is requested to receive, possess, store, and package for return to the vendor unieradiated fuel assemblies in the amount necessary for operation of Braidwood Station Unit 2.
Braidwood Station is being constructed by the Commonwealth Edison Company at a site located in northeastern Illinois, 3 miles southwest of the Kankakee River, 20 miles south-southwest of the town of Joliet, and 60 miles southwest of Chicago, Illinois.
The station is within Reed Township, Will County, Illinois. Construction Permit No. CPPR-133 (Docket No. 50-457) was issued on December 31, 1975 for Braidwood Unit 2.
1.1.2 Braidwood Station will be fueled with Westinghouse 17x17 Optimized PWR Fuel Assemblies.
Each fuel assembly consists of a bottom nozzle, a top nozzle, 24 control rod and insert guide tubes, one instrument tube, 8 grid assemblies and 264 fuel rods.
Refer to Figures 1.1-1 and 1.1-2 (Byron /Braidwood FSAD Figuree 4.2-1 and 4.2-2 respectively).
The fuel rods consist of uranium dioxide (UO ) ceramic pellets 2
contained in Zircaloy-4 tubing which is plugged with Zircaloy-4 end plugs and seal welded to encapsulate the fuel. This Zircaloy-4 cladding has a nominal thickness of 0.0225 inches and a nominal outside diameter of 0.360 inches.
The fuel pellets are right circular cylinders, with a nominal diameter of 0.3088 inches and a nominal height of 0.507 inches, consisting of slightly enriched UO2 powder which has been cold pressed and sintered to 95% of its theoret.ical density. The ends of each pe13et are dished slightly to allow greater axial expansion at the center of the pellets, and shifting of the fuel within the cladding during handling or shipping prior to core loading is prevented by a stainless steel helical spring which bears on top of the pellet stack. All fuel rods are internally pressurized'with helium during the welding of end plugs, in order to minimize compressive cladding stresses and prevent cladding flattening due to Reactor Coolant System operating pressures.
Refer to Figure 1.1-3 (Byron /Braidwood FSAR Figure 4.2-3).
The 264 fuel rods are positioned in a 17x17 square array with a rod pitch of 0.496 inches. Table 1.1.2 summarizes fuel assembly and fuel rod data.
The bottom nozzle serves as the bottom structural element of the fuel assembly and directs the coolant flow distribution to the assembly.
The 8.424 inch (nominal) square nozzle is fabricated from Type 304 Stainless Steel and consists of a perforated plate and four angle legs with bearing plates. The legs form a plenum for the inlet coolant flow to the fuel assembly, and the perforated plate prevents accidental downward ejection of fuel rods from the fuel assembly since penetrations in the plate are positioned between the rows of fuel rods.
The bottom nozzle is fastened to the fuel assembly guide tubes by locked screws which penetrate through the nozzle and mate with a threaded plug in each guide tube.
Indexing and positioning of the fuel assembly is controlled by alignment holes in two diagonally 1
0661m(041387) 0327A
A Special Nuclear Material License is requested which authorizes the receipt, possession, inspection, storage and packaging for return to the vendor unieradiated fuel assemblies made of uranium enriched in the U-235 isotope. All fuel assemblies are for use at the Braidwood Nuclear Generating Station, Unit 2.
Regulatory Guide 3.15 (Task CE 219-4), " Standard Format and Content of License Applications for Storage Only of Unieradiated Power Reactor
{
Fuel and Associated Radioactive Material", Revision 1, April 1983 (Attachment 1
A) was used to prepare the following application. All reference material used i
to support the application is included as a separate attachment.
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BYRON /BRAIDWOOD STATIONS FINAL SAFETY ANALYSIS REPORT Figure 4.2-1, 17 x 17 Optimized Fuel Assembly Cross Section 2
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SPECIFIC DIMENSIONS DEPEND ON DESIGN VARI ABLES SUCH AS PRE PRESEUR12ATION, POWER HISTORY, AND DISCHARGE BURNUP BYRON /BRAIDWOOD STATIONS-FINAL SAFETY ANALYSTS REPORT Figure 4,2-3.
Fuel Rod Schematic i
0%E;94 013787 D
TABLE 1.1.2 Braidwood Nuclear Generating Station Unit 2 General Fuel Data Fuel Assembly Data i
Overall Length 159.8 inches i
Nominal Active Fuel Length 144 inches I
Fuel Rod Pitch 0.496 inches 1
Rod Array 17 x 17 l
Rods Per Assembly 264 i
Control Rod and Insert Guide Tubes Per Assembly 24 j
l Instrumentation Tube Per Assembly 1
Fuel Rod Data Inches I
Outside Diameter 0.360 Cladding Inside Diameter i
0.315 l
Cladding Thickness 0.0225 i
Fuel Pellet Density (% theoretical) 95.0 Fuel Pellet Diameter 0.3088 i
Fuel Pellet Length 0.507 5
0661m(041387) 0327A
opposite bearing plates which mate with locating pins in the reactor
. vessel's lower core plate.
The top nozzle functions as the upper structural element of the fuel assembly and provides a partial protective housing for the rod cluster control assembly (RCCA) or other fuel assembly inserts. It consists of an adapter plate, enclosure, top plate, and pads. Holddown springs and their attachment bolts made of Inconel-718 are mounted on the top nozzle which is made of Type 304 Stainless Steel. The 8.404 inch (nominal) square adapter plate has round holes and semi-circular ended slots positioned between rows of fuel rods to permit the flow of coolant-upward through the top nozzle, and to prevent the upward ejection cf rods from the fuel assembly.
Other round holes.are-
. provider n,' accept sleeves which are welded to the~ adapter plate and mechanically attached to the guide tubes.
The enclosure is a box-like structure which sets the distance between the adapter plate and the top plate. The top plate has a large square hole in the center to permit access for the control rods.ano other insertr.
Four holddown springs are mounted on the top plate and are.
bolted in place at two diagonally opposite corners. On the other two corners, integral pads are positioned which contain alignment holes.
These alignment holes are on the same corners as the bottom nozzle alignment holes and they mate with ?ocating pins in the reactor vessel's upper core plate.
The 24 guide tubes are structural members which also provide channels for neutron absorber rods, burnable poison rods, primary and secondary neutron sources, or thimble plug assemblies. Each tube is fabricated
]
from Zircaloy-4 tubing having two different diameters. The tube diameter at the top section provides the area necessary to permit rapid control rod insertion during a reactor trip.
The lower portion of the guide tube is swaged to a smaller diaueter to produce a dashpot i
action near tha and of the control rod travel during normal trip operation. Holes are provided on the tube above the dashpot to reduce the rod drop time.
The dashpot is closed at the bottom by means of an
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end plug which has a small flow hole to avoid fluid stagnation in the dashpot volume during normal operation. The top end of the guide tube is fastened to a tubular sleeve by three expansion lobes.
The sleeve fits into and is welded to the top nozzie adapter plate. The lower end of the guide tube is fitted with an end plug which is then fastened into the bottom nozzle by a locked screw.
Refer to Figures i
1.1-4 and 1.1-5'(Byron /Braidwood FSAR Figures 4.2-6 and 4.2-7 respectively).
The central instrumentation tube of each fuel assembly is constrained by seating in counterbores in each nozzle. This tube is made of Zircaloy-4 and is a constant diameter.
It guides'the.incore thimbles which are sealed at the reactor end and at the seal table.
The incore thimbles thus provide an extension of the primary coolant system pressure boundary and allow the incore neutron fission detectors to travel up the center of 58 selected fuel essemblies.
6 0661m(041387) 0327A
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0661rn(011787 )
The fuel rods are supported at 8 intervals along their length by two types of grid assemblies which maintain the lateral spacing between the rods.
Each fuel rod is supported within each grid by the combination of support dimples and springs.
Both types consist of individual slotted straps interlocked in an egg-crate arrangement.
The middle six grid assemblies are made of Zircaloy-4 and their internal straps include mixing vanes which project into the coolant stream and promote mixing of the coolant. The two end grid assemblies are made of Inconel-718 and do not have mixing vanes on the internal straps.
Each grid is fastened to the guide tubes to create an integrated structure.
The middle six Zircaloy grids have Zircaloy sleeves welded to them.
An expanding tool is inserted into the inner diameter of the guide tube at the elevation of the sleeve and forces the tube and sleeve outward to a predetermined diameter, thus joining the two i
components. The top grids have stainless steel sleeves which are joined to the guide tubes by expanding the two members and then i
welding the top ends of the sleeves to the top nozzle adaptor plates.
A stainless steel insert is spotwelded to the bottom grid and captured j
between the guide tube end plug and the bottom nozzle with a stainless steel screw.
1.1.3 The maximum U-235 enrichment requested is 3.5 w/o with a manufacturing tolerance not to exceed 1 0.05 W/o.
The maximum quantity of U-235 requested per assembly is 14.81 kg based on 3.5-W/o enrichment. The total estimated weight per assembly is about 1358 lbs. (616.0 kg) which breaks down as f'llows:
o Material Weight lbs(kg)
Uranium dioxide 1050 (476.3)
Zircaloy 266 (120.7)
Inconel 5
(2.3)
Stainless Steel 37 (16.8)
Total per assembly 1358 lbs. (616.1 kg)
There are no quantities of U-233, plutonium, depleted uranium or thorium per assembly.
1.1.4 A license is requested to possess and store a total of 198 new fuel assemblies of various U-235 enrichments.
The total weight of U-235 requested is 2186.07 kg which is broken down by fuel enrichment in the following table:
9 0661m(041387) 0327A
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Weight'of U-235'in Puel Assemblies (Enrichments are weicht'% U-235)
' Weight (kg)
Description Unit 2
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1.
193' initial core assemblies a.
65 Region 1 8 2.1 W/o
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64 Region 2 6 2.6 W/o 703.87 c.
64 Region 3 6 3.1 "/o 839.23' 2.
5 initial core replacement assemblies 65.57 6 3.1 "/o enrichment 2186.07-Total weight of U-235' 2186.07 kg.
The above quantities of U-235 were calculated based upon an average mass of 423 kg U per assembly. Nominal enrichments of 2.1, 2.6, or 3.1 "/o were'used and manufacturing tolerances were not considered in the calculation. More precise enrichment values.will be available' when the fuel assemblies are actually fabricated. There is no U-233, plutonium, depleted uranium or thorium contained in the fuel assemblies for which a license is requested.
)
w 10 0661m(041387) 0327A
1.2 Storage Conditions 1.2.1 Fuel assemblies will be inspected and stored in the Fuel Handling Building.
Attached Figures 1.2-1, 1.2-2, 1.2-3 and 1,2-4 show these areas.
1.2.2 New fuel will be stored dry in the new fuel racks and either wet or dry in the spent fuel pool storage racks.
1.2.3 During the construction phase when new fuel is stored in the spent fuel pit, the adverse effects of dropping a construction load being carried by the Fuel Handling Building Crane can potentially threaten the safety of storage. During this phase, the Fuel Handling Building Crane will be handling loads which are below 15 tons.
The majority of the loads carried will include pipe,-valves, scaffolding, trash, and pipe supports.
A temporary heavy load exclusion area, over the Braidwood new fuel storage area of the spent fuel pit, will be defined to preclude the possibility of dropping a heavy load on fuel stored in the spent fuel i
racks.
Electrical interlocks will prevent auxiliary hook motion in the area west of a line which is 15 feet east of the spert fuel pool east wall and north of column 19 in the Fuel Handling Building. No heavy loads will be transported within the temporary exclusion area.
All loads will be transported along the designated heavy load removal path shown on Figure 1.2-1 which is Byron /Braidwood print M-27 sheet 7.
All fuel stored in new or spent fuel racks will be covered with steel cover plates or two layers of aluminum respectively.
Thus the heavy load exclusion area and the fuel rack covers combine to minimize the potential effects of construction activities on the safety of storage.
1.2.4 The Fuel Handling Building where fuel will be stored is located adjacent to the auxiliary building between the containment structures, Figure 1,2-1.
It is a reinforced concrete structure up to grade, except in the fuel pit area where reinforced concrete is continued up to the mezzanine level. The walls below grade bear on a reinforced concrete mat foundation.
The portion of building above grade has a structural steel frame with concrete slab on metal deck.
The exterior walls are concrete for radiation shielding and missile protection.
The interior walls are of either concrete or concrete block construction.
Fuel access is at grade level to permit use of an overhead crane for handling the fuel.
New fuel is stored in racks located in a reinforced concrete vault as shown in Figure 1,2-1.
The new fuel storage vault provides dry storage for 132 fuel assemblies which is approximately two-thirds of a core.
Each rack consists of lockable vertical cells of stainless 11 0661m(041387) 0327A
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N LEVELNG LEG q q gg$g a MOLC DOWN BOLT BYRON /BR AIDWOOD ST ATIONS l
FIN AL S AFETY AN ALYSIS REPORT FIGURE 9,1-3 15 TYPICAL SPINT FUEL RACK ISOMETRIC 06Eh.(0117E7)
'J n,.
]
steel construction. Three groups of 44 cells each,'in' rows of two, at
{
a center-to-center spacing of 21 inches are attached to the support structure. The support structure is bolted and/or welded to embedded studs in the concrete walls and columns of the storage vault.
Fuel
]
assemblies rest on self-leveling support plates at the bottom of each j
cell, which in turn sit on Douglas Fir pads. All surfaces that contact fuel assemblies are austenitic stainless steel and the j
supporting structure is made of high strength carbon steel. The racks rest on an 18-inch floor slab with 2-foot thick reinforced concrete walls and are designed to withstand normal operating loads as well'as design-basis seismic loads. Access'to the vault is through openings in the floor at elevation 426 feet.
The spent fuel storage racks provide a place in the spent fuel storage-pool for storing the spent fuel discharged from the reactor and for i
temporarily storing new fuel for initial core loading.
There are 1050 top entry racks designed to ma'intain the spent fuel in'a space geometry that precludes the possibility of criticality under both normal and abnormal conditions. The spent fuel storage racks can also be used for' dry storage of initial core fuel with up to 3.20 w/o U-235 enrichment in a checkerboard pattern only or wet storage of initial core fuel with up to 3.20 w/o U-235 enrichment in borated water of at least 2000 ppm boron concentration, prior to the initial core loading. The location of the spent fuel pool within the plant is shown in Figure 1.2-1.
The spent fuel storage racks (shown in Figure 1.2-4) consist of square-stainless steel tubes of 1/8-inch thick type'304 Authentic Stainless Steel. They are held at 14 inches center-to-center by 1/8-inch thick type 304 Stainless Steel plates.
The plates are welded to the sides of the square storage tubes at four elevations. The tubes are flared-at the top to permit easy storage and retrieval of the stored fuel assemblies, and to be compatible with the fuel handling equipment.
)
Two different rack arrays (6x5 and 6x6) are utilized to maximize use of the available fuel storage space in the pool. The rack structure is welded to an elevated base which is a system of welded beams and stiffeners. The base serves to support the weight of the fuel assemblies and to distribute the load on the pool floor.
The base contains an opening at each fuel assembly storage location which accommodates the fuel assembly lower nozzle. The storage cells are designed to provide lateral support for stored assemblies of the Westinghouse 17x17 array design.
To prevent movement of the racks, particularly from seismic excitation, the racks are bolted to embedment plates in the floor-of the spent fuel pool. The racks are raised off the floor on adjustable legs to allow natural circulation flow with spent fuel stored in the racks. The legs are adjusted after installation to level'the racks and distribute the load to all legs.
l 16 0661m(041387) 0327A
I 1
1 l
The fuel handling building which encloses the spent-fuel. pool, new fuel storage vault, and port ions of the fuel transfer system is a Seismic Category I building.
It is therefore capable:of withstanding
_j the design-basis earthquakes, tornados, and missiles and is not J
subject to flooding. The new fuel storage racks'are designed to preclude storage of a fuel assembly other than where intended and to withstand the uplift force which could occur due-to a fuel assembly 1
hanging up during lifting.
These racks can also withstand the impact
]
load of a dropped fuel assembly.
Spent fuel racks were also' designed to withstand either a dropped fuel assembly or the upward force of a stuck assembly _ without loss of structural integrity.
1 The following fuel handling equipment and components may be used to handle, inspect, and place fuel into storage until initial core loading:
1.
Fork lift truck - larger than 6700 pounds lifting capacity.
2.
Fuel handling building crane - a motorized overhead crane to I
service the fuel handling building, rated capacity is 125 tons with one 15 ton auxiliary hoist.
3.
New fuel handling tool - short handled tool used to transfer new.
fuel assemblies from the unloading area to the new fuel elevator 4
or new fuel storage via the fuel handling building crane.
For-receipt and storage of first core fuel, the new fuel handling tool may be used to directly insert new fuel into the spent fuel _
racks. The tool has been preop'erationally tested at 125% of the weight of one fuel assembly.
4.
Spent fuel handling tool - a long handled tool used in the spent,
1 fuel pool with the spent fuel pit bridge crane.
T,he tool has been preoperationally tested at 125% of the weight of one fuel assembly.
5.
New fuel elevator - the new fuel elevator, load rating 2000 pounds, consists of a box shaped elevator assembly with its top end open and sized to house one fuel assembly. The new fuel elevator is used to lower a new fuel assembly to the bottom of the fuel storage area where it is transported to the storage racks by the spent fuel pit bridge crane.
6.
Spent fuel pit bridge crane - a' motor operated traveling bridge crane with overhead hoist to position personnel and tools over the spent fuel pool for refueling.
The bridge will span the pool.
Fuel assemblies are moved between storage rack positions with the spent fuel handling tool suspended.from the hoist and manipulated by an operator on the bridge walkway.
The hoi'st capacity willf be 4000 pounds.
17 0661m(041387) 0327A
1 i
1.2.5 The fire protection measures for the Fuel Handling Building consist of separation of the area from the rest of the plant by a fire barrier.
Fire dampers are provided in the fire wall.
Fire dampers are fusible link type which will be closed when the temperature exceeds 165'F.
There are no safety-related cable trays in this area. All cables in.
the area not in motor-control cubicles are in conduit. The new fuel storage racks are designed to preclude criticality when subjected to water spray. Nonfogging type nozzles are used for the manual hose
]
stations located in this area to preclude the optimum moderation i
condition for the new fuel storage racks.
Refer to Byron /Braidwood Fire Protection Report, Section 2.3.12 (Attachment B).
Portable fire extinguishers and four manual hose stations constitute the fire protection equipment. Two portable fire extinguishers and two manual hose stations are located on elevation 426 feet, and two manual hose stations on elevation 401 feet.
Similar equipment can be brought in from the Auxiliary Building at elevation 426 feet, if needed.
Ionization detectors are provided for elevation 401 feet which alarm and annunciate locally and in the Control Room. The large area above elevation 426 feet is provided with ultrat tolet detectors, which will be tr. ore effective than ionization detectors for this area due to the large room volume.
These also alarm and annunciate in the Control Room.
1.2.6 The Security Plan for Special Nuclear, Material at Braidwood Station will be implemented during fuel receipt,~ handling, and storage.
The following are areas where new fuel may reside for varying periods of time prior to reactor core loading:
1.
The area immediately outside the fuel handling building.
2.
The following areas of the fuel handling building:
Unioading area of the fuel handling building trackway a.
b.
New fuel unloading area (for removal from shipping containers) c.
New fuel storage racks d.
Spent fuel storage pool 18 0661m(041387) 0327A
During plant construction, whenever fuel is being received, handled or stored-in any of these areas, access to that area will be controlled.
The controlled access area will,be' defined by using a locked physical
~
barrier or by watchmen stationed at each functional entrance to the area.1 Unauthorized access will be prevented by manning the personnel entry points of these areas to limit entry to those individuals identified as having a need to enter.
The' physical barriers will be patrolled by watchmen to verify continued integrity and any irregularities shall be reported immediately to the Shift. Supervisor.
An access authorization list will be maintained which lists all individuals-having a need to work within the controlled 1access areas.
1.3 Physical Protection No'U-233 or plutonium is to be possessed under the license and the quantity of U-235 to be possessed is classified as "special nuclear material of. low strategic significance". Commonwealth Edison Company will comply with all applicable sections of.10 CFR Part 73, '" Physical Protection of Plants and Materials", particularly paragraph 73.67(f),
Fixed Site Requirements for Special Nuclear Material of Low Strategic Significance, as outlined in the " Security Plan for Special Nuclear Material Security for' Commonwealth Edison Company,'Braidwood Station".
1.4 Transfer of Special Nuclear Material 1.4.1 Westinghouse, the fuel fabricator, is responsible for.the shipment of fuel to Braidwood Station.
1.4.2 Any unieradiated fuel to be returned to Westinghouse will be packaged and shipped in the containers originally used te ship fuel to Braidwood Station.
1.4.3 The applicable provisions of 10 CFR Part 70 related to the control and accounting practices of special nuclear material are complied with using Commonwealth Edison Company's (CECO) Nuclear Procedures Nos. 1 through 8 (Attachment C).
These Nuclear Procedures clearly define the organization, responsibilities, and facilities comprising CDCo's Nuclear Material Control System, which' provide a method for detecting any loss or unauthorized diversion of' nuclear material.
Am index of specific nuclear procedures is shown in Table 1.4.3.
A few key positions in implementing the CECO Nuclear Procedures and their responsibilities are discussed below. The Nuclear Materials Safeguards Manager (NMS Manager) will have overall responsibility for surveillance and control of nuclear material. His specific. duties include preparing and filing Nuclear Material Transaction Reports, Nuclear Material Balance Reports and Physical Inventory Listings as may be required.
19 0661m(041387) 0327A-
The Nuclear Materials Custodian (NMC) shall be responsible for:
(1) inventorying and maintaining records of the movements and locations of all special nuclear material within the Item Control Areas (ICAs)'
assigned to him, (2) ensuring that all fuel vaults are properly secured after each inventory or other use, and (3) preparing all necessary reports to verify all receipts of nuclear fuel delivered to the station, all shipments from the station, and~all transfers within the station-site.
The Nuclear Fuel Services Manager shall be responsible for the operation of the Nuclear Component Accountability System (NCAS) which shall include the recordkeeping for all nuclear fuel assemblies, the computation of all nuclear fuel burnup, and the total element and isotopic composition of all nuclear material in nuclear fuel assemblies in the possession of Commonwealth Edison Company.
He shall document and retain for NRC audit purposes the methods used in the computation and compilation of nuclear fuel data for inclusion in the NCAS.
The Director of Auditing shall be responsible for observing and performing audit tests of physical inventories of nuclear fuel assemblies in a reactor upon completion of and immediately prior to installing the reactor head after initial fueling, refueling, or rearrangement of fuel, and of reportable quantities of Special Nuclear Material in Fuel Pools, Fuel Vaults, and other storage locations when such audited inventories are required.
He shall also be responsible for performing annual audits to ensure compliance with the Company's Nuclear Procedures and such other audits as requested by the NMS Manager.
j 20 0661m(041387) 0327A
a p
1
'j Table 1.4.3
_j INDEX OF NUCLEAR PROCEDURES l
N.P. NO.
Title 1
Organization, Responsibilities, and Facilities for Controlling and Safeguarding Nuclear Material a
2 Safeguarding and Controlling Shipments of Nuclear Fuel 3
Safeguarding and Controlling Movements of Nuclear Fuel Within a Station I
4 Physical Inventories of Nuclear Fuel. and Other Special Nuclear Material (SNM) Items 5
Reporting Nuclear Fuel /SNM Discrepancies and Unprotected Conditions y) i 6
Control of Nuclear Licenses and Certificates of Registration
}
7 Reporting Inventories of Nuclear Materials to the NRC 5
8 Receipt and Shipment of Non-Fuel Items Containing Non-Exempt
)
Quantities of Special'-Nuclear Material (SNM)
H I
21 0661m(041387) 0327A
1.5 Financial Protection and Indemnity 1.5.1 Pursuant to 10 CFR 140.15, Commonwealth Edison will furnish proof of financial protection in the amount of $1,000,000' in the form of a Nuclear Energy Liability. Insurance Association Policy application.
Copies of the policy, when issued, will be sent to the Office of State Programs.
An indemnity agreement will be executed with the Commission effective on the date of issuance of the Braidwood Station Special Nuclear Material License for fuel storage. The annual. indemnity fee applicable to preoperational fuel storage will be paid by Commonwealth Edison.
1.5.2 Not Applicable 1.5.3 Not Applicable l
Chapter 2 Health and Safety 2.1 Radiation Control 2.1.1 Key positions having radiation-safety responsibilities are the Station Manager, Radiation / Chemistry Supervisor, and Station Health Physicist. Minimum qualifications for these positions are specified in Regulatory Guide 1.8, " Personnel Selection and Training" and/or_
ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel." The minimum technical qualifications for the Station Manager shall be in accordance with Section 4.2.1, " Plant Manager," of ANSI N18.1-1971. The minimum technical qualifications for the.
Radiation / Chemistry Supervisor shall be in accordance with Sections 4.4.3 and 4.4.4, " Radiochemistry" and " Radiation Protection", of ANSI N18.1-1971. These qualifications may alternately be met by technical i
personnel reporting to the Radiation / Chemistry Supervisor.
The minimum technical qualifications for the Radiation / Chemistry Supervisor or the Station Health Physicist shall be in accordance with the requirements for Radiation Protection Manager of Regulatory Guide 1.8, September 1975.
2.1.2 The Station Manager exercises overall managerial and supervisory responsibility for the safe operation of the plant and its equipment.
He is responsible for compliance with the station's NRC licenses, government regulations, ASME code requirements, and the company quality assurance program. The Station Radiation / Chemistry Supervisor is responsible for the Station chemistry and radiation protection programs and for coordinating the emergency planning and' environmental monitoring programs. He has access to the Station Manager on matters of radiation protection during unusual or abnormal radiological conditions. The Station Health Physicist is responsible for daily.
health physics activities. He reviews health physics surveys', dose data, internal deposition data, meteorological data and environmental data. He participates in the health physics aspects of emergency planning and environmental monitoring activities as assigned. He also is capable of assuming the duties of the Radiation / Chemistry Supervisor when assigned.
22 0661m(041387) 0327A
l 1
2.1.3 The~ qualifications and experience of key personne responsible for radiation safety are provided in the resumes:incladed in' Attachment D.
t i
2.1. 4 -
Surface contamination is Metected through: the use of Geiger-Mueller (GM) type instruments in conjunction with a pancake type probe..A 100 cm2 smear is taken'and evaluated using the'GM.- If an area is i
determined to be contaminated,'the area is roped off, posted and decontaminated when practical.
Decontamination is performed by
. wiping', sweeping or mopping the area until the contamination level-(above background) does not exceed 1000 dpn/100 cm2 removable, for-beta and gamma emitting isotopes.
If decontamination cannot be achieved, the area is posted and routinely surveyed.
1 Personnel contamination is detected through the use of portal-monitors, friskers,'or Geiger-Mueller type instruments.
A survey is j
performed on the body.to determine the exact location of J
contamination. Decontamination is performed by washing with varying.
strengths of detergent.
If washing does nc>t remove the contamination, a 4% Potassium permanganate solution may be applied to the skin followed by a Sodium Bisulfide solution.
'1 i
2.1.5 Radiological protection instruments will be calibrated on a quarterly.
basis.
Dose rate instruments will be calibrated by a 400 Ci J.L.
Sheperd Calibration Unit, Model No.'89.
A performance test will be
-]
performed prior to each use of the instrument and repeated every four hours during continuous use.. Count rate meters will be electronically i
calibrated as well as source checked on a quarterly basis.
A l
performance test will be performed prior to.each use of these i
instruments and repeated every four hours'during continuous use.
l 2.1.6 Braidwood Station plant specific Radiation Protection Procedures and the Commonwealth Edison Company Radiation Protection Standards i
(Attachment E) will be utilized in complying with applicable sections of 10 CFR Part 20, " Standards for Protection Against Radiation".
It is the policy of Commonwealth Edison Company to maintain occupational radiation exposures as low as is reasonably achievable (ALARA),
consistent with station construction, maintenance, and operational requirements, and with economic and social considerations being taken into account.-
/
2.1.7 The radioactive waste system provides equipment necessary to collect, process, and prepare for the disposal of radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation.
After collection, depending on chemical' composition liquid wastes are evaporated, demineralized and/or filtered.
The treated water may be recycled for use in the plant or may,be' discharged at concentrations within the limits of 10 CFR 20.
The evaporator concentrates and the spent demineralized resins are solidified for ultimate' disposal;in'an authorized location. Gaseous. wastes are collected from the waste gas header. Discharge of the gaseous wastes to the enviornment is controlled to ensure that the offsite dose ~is as low as reasonably achieveable. (ALARA). All applicable station drumming and waste' handling procedures, to minimize personnel exposure, will be followed I
in accordance with CECO's ALARA policy.
23 0661m(041387)/0327A
2.2.
Nuclear Criticality Safety 2.2.1 Key positions having nuclear criticality safety and fuel handling responsibilities aro the Station Manager, Operating Engineer, Fuel Handling Foreman, and Technical Staff Supervisor.
Minimum qualifications for these positions are specified in ANSI N18.1-1971,
" Selection and Training of Nuclear Power plant Personnel." The minimum technical qualifications for the Station Manager shall be in accordance with Section 4.2.1, " Plant Manager," of ANSI N18.1-1971.
The minimum technical qualifications for the Operating Engineer shall be in accordance with Section 4.3.2, " Supervisors Not Requiring AEC Licenses", of ANSI N18.1-1971.
The minimum technical qualifications for the the Fuel Handling Foreman shall be in accordance with Section 4.3.1, " Supervisors Requiring AEC Licenses," of ANSI N18.1-1971.
The j
minimum technical qualifications for the Technical Staff Supervisor shall be in accordance with Section 4.6.1, " Engineer in Charge," of ANSI N18.1-1971.
l 2.2.2 The Station Manager exercises overall managerial and supervisory j
responsibility for the safe operation of the plant and its equipment.
)
He is responsible for compliance with the station's NRC licenses, government regulations, AEME code requirements, and the company
{
quality assurance program. The Operating Engineer is responsible for the operation of the mechanical and electrical equipment and certain common plant systems, such as radioactive waste processing and fuel handling. He is also responsible for authorizing functional acceptance tests to be conducted by Operation and/or Technical Staff i
personnel. New fuel is handled under the supervision of the Fuel i
Handling. Foreman. He is responsible for developing and implementing procedures involving the control and handling of nuclear fuel.
The Technical Staff Supervisor provides technical support for plant
{
operations, refueling, maintenance, and modification and evaluates process data, equipment performance, and adequacy of station procedures. He also has the responsibility and authority for implementation of the onsite review function.
Proceduress and changes thereto, for the control and handling of nuclear fuel shall be reviewed and approved by an Operating Engineer and the Technical Staff Supervisor. At least one individual who reviews and approves fuel handling procedures will hold a SRO License for Braidwood Station.
Nuclear Component Transfer Lists will only require Technical Staff Supervisor approval unless core alterations are involved, in which case Services Superintendent or Assistant Superintendent-Technical Services approval is also needed.
24 0661m(041387) 0327A
2.2.3
. Fuel assemblies will be received on site in Westinghouse model RCC-1 or RCC-3. shipping containers _ presently licensed under NRC license
- USA /5450/AF.
This shipping container is a reusable metal container designed for shock and vibration isolation, and' sufficient leak tightness to protect fuel assemblies from damage during normal handling and shipping at temperatures from -40*F to 150*F.
Each container may contain one or two fuel assemblies with or without inserts such as rod control cluster assemblies, burnable poison rod assemblies, thimble plugs, or source assemblies.
Shipments are made via flatbed truck with a maximum of 14 assemblies (7 shipping containers) per truck.
The weight of a fully loaded ' shipping container 'will not exceed 6700 pounds.
Normally fuel assemblies will not be stored in their shipping containers for a significant length of time, but loaded containers may be stcred temporarily in the receiving j
area of the Fuel Handling Building until fuel inspection is
{
completed. Westinghouse's license allows the shipment of up to 60 j
containers at one time, and they have indicated that the number of loaded shipping containers that may be located together is not a criticality concern. Therefore, no special controls are required to ensure the nuclear criticality safety of.the temporary storage of loaded shipping containers.
2.2.4 After a shipment of fuel assemblies is received it will be unpackaged, inspected and stored in either the new or spent fuel racks.
For the following description of the new fuel racks refer to Figure 1.2-2 (Byron /Braidwood FSAR Figure 9.1-1).
132 new fuel storage cells are laid out in 3 groups of 44 storage cells each.
Each group of 44 storage cells is made up of 2 rows of 22 storage cells each The lattice spacing in either direction for storage cells within each group of 44 cells is 21.00" typical.
The lattice spacing between the closest cells of adjacent groups of 44 cells is greater than 65",
Each Type 304 Stainless Steel cell in the new fuel storage racks have the following specifications:
the inner dimension is 9.00 (+0.12,
-0.00) inches square at all cross sections; the. wall thickness is 0.125 1 0.010 inches; and the outer dimension is therefore a nominal 9.25 inches square.
New fuel will be held off the floor of the new fuel vault by resting on self-leveling support plates at the bottom of each cell which, in turn, sit on top of nominal 2.00 inch Douglas Fir pads.
The east side of the racks measures nominally 36.5" from the east vault wall to the east edge of the fuel cells.
The west side of the racks measures nominally 31.5" from the west vault wall to the_ west edge of the fuel cells.
There is at least 69" to the edge of the closest fuel cell as measured from the south end of the racks to the south vault wall.
There is at least 51" to the edge of the closest fuel cell'as measured 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.
25 0661m(041387) 0327A
8 Refer to Figure 1.2-3 (Byron /Braidwood FSAR Figure 9.1-2) for the following description of the spent fuel racks.
1050 spent fuel storage cells are laid out in 20 6x6 racks and 11 6x5 racks.
An additional 10 cell failed fuel storage rack is located in the northeast corner of the spent fuel pit, having the same external size as a 6x5 rack.
The lattice spacing in either direction for storage cells within each 6x5 or 6x6 rack is 14.00" typical.
Spacing between 3
racks in the north-south direction is 1.50" typical (to obtain a 14.5" typical pitch for perimeter fuel cells).
Spacing between racks in the i
east-west direction is 7.00" typical (to obtain a 20.00" typical pitch for perimeter fuel cells).
The distance from the bottom of the fuel assemblies placed in the spent fuel racks to the floor of the spent fuel pit will exceed 13.00".
The east wall of the spent fuel pit is located nominally 28.50" from the east side of the racks closest to it.
The west wall of the spent fuel pit is located nominally 15.50" from the west side of the racks closest to it.
The west half of the north wall of the spent fuel pit is located nominally 13.50" from the north side of the
)
racks closest to it.
The east half of the north wall of the spent fuel pit is located nominally 20.50" from the north side of the racks closest to it.
The south wall of the spent fuel pit is located nominally 13.50" from the south side of the racks closest to it.
No l
neutron poisons are used in the fabrication of the spent fuel storage racks.
New fue'l assemblies may be stored either wet or dry in the spent fuel rack positions. A checkerboard loading pattern will be used for dry i
storage, with the four nearest adjacent positions to a loaded cell empty.
For those spent fuel rack storage cell locations on the perimeter of the spent fuel pit, the nearest adjacent positions ~are to I
remain empty (there may be 2 or 3 nearest adjacent positions).
Two l
independent persons will verify placement of the new fuel assemblies in the spent fuel pit storage rack locations for dry storage to ensure the checkerboard pattern is maintained. Additionally, during dry storage, a new fuel assembly in transport will not be allowed to pass over a storage cell that is immediately adjacent to a cell that a,1 ready contains a fuel assembly.
For wet storage, a normal loading pattern may be used provided the water is of at least 2000 ppm boron concentration.
The integrity of new and spent fuel storage racks is discussed in Section 1.2.4 of this application.
Controls which limit the supply of q
combustible materials in fuel storage areas will be in accordance with descriptions given in the Byron /Braidwood Fire protection Report, Sections / Subsections 2.3.12.1, 3.2(i), and 3.7(1), (Attachment B).
Douglas Fir pads are used at the base of fuel cells in the new fuel vault and miscellaneous lumber is temporarily used for construction related activities in other areas of the Fuel Handling Building.
No special fire protection features are considered necessary.
I i
I i
l 26 0661m(041387)'
j 0327A
1 The Braidwood Station, Safety Evaluation Report, NUREG-1002, sections 9.1.1 and 9.1.~2 (these sections are the same as in the Byron Station, SER, NUREG-0876, Attachment F) discusses new fuel storage in the new fuel vault and spent fuel racks, and concludes that both storage facilities are acceptable.
2.2.5 The nuclear criticality safety of our new fuel assembly storage is based on values that bound'the maximum U-235 enrichments. The maximum U-235 enrichment requested is 3.50 W/o with a manufacturing.
tolerance not to exceed 1 0.05 W/o.
The new fuel storage racks were analyzed for dry storage of 4.00 W/o enriched fuel assemblies with no limitations placed on the loading pattern.
The new fuel storage racks were also analyzed for wet storage of 4.00 W/o enriched new fuel assemblies with no limitations placed on the loading pattern.
The spent fuel storage racks were analyzed for dry storage of 3.20 +
0.015 W/o enriched fuel assemblies in a checkerboard pattern only, prior to the initial core loading. The spent fuel storage racks were also analyzed for wet storage of 4.00 W/o enriched new fuel assemblies with no limitations placed on the loading pattern.
2.2.6 Nuclear criticality safety is not based on the. reactivity effects of s
neutron absorber materials in the racks since no neutron poisons are used in the fabrication of the new and spent fuel storage racks.
_ 2.2.7 Nuclear criticality safety is not based on moderation control since k gg was shown to be within ANSI requirements when the optimum e
moderation condition was assumed in.the safety analysis.
K gg_will e
be less than 0.98 (in accordance with ANSI N18.2-1973) with fuel of a-maximum enrichment of 4.00 W/o U-235 stored in the new fuel vault, assuming optimum neutron moderation conditions (dry or fogged).
K gg will be less than 0.95 (in accordance with ANSI N210-1976) with e
fuel of a maximum enrichment of 4.00 W/o U-235 in place in the new fuel vault, assuming the stored assemblies are completely submerged in unborated water at a conservative water temperature with'no credit for neutron poison in the fuel assembly.
The optimum moderation' condition exists when the new fuel racks are covered with clean water at a uniform density of 0.018 gm/cc.
Provisions have been made for draining water that may accumulate around the stored fuel assemblies as a leakage path exists around the edges of the self-leveling support plates at the bottom of the individual storage cells.
i 27 0661m(041387) 0327A
I For the spent fuel racks, k gg will be less.than 0.95 for the normal e
wet condition (in accordance with ANSI N210-1976) with fresh fuel of up to 4.00 W/o U-235, a conservative water temperature, no credit for fixed neutron poison in the fuel assemblies or soluble neutron poison in the pool water, and no credit for structural material other than the stainless steel cans.
The design of the spent fuel storage racks also allows a subcritical effective multiplication factor for dry storage of new fuel with 3.20 W/o enrichment in a checkerboard loading pattern:
k gg less than 0.98 for the optimum moderation e
condition and k gg less than 0.95 for the flooded condition in e
accordance with ANSI N210-1976. The optimum moderation condition i
exists when the storage racks are covered with clean water at a uniform density of 0.05 gm/ce.
The flooded condition exists when the storage racks are covered with cold, clean, unborated water at a full density.
2.2.8 The criticality analyses of the new fuel racks for the optimum moderation and flooded conditions inc1r ad several conservative assumptions as well as the effects of uncertainties in the calculation j
method, and geometric and material variations of the fuel storage rack. The following conservative assumptions were used in the calculation:
Fresh fuel of 4.00 W/o U-235 enrichment - Initially the maximum a.
enrichment will be lower than this, but it could approach thiz enrichment if an 18-month fuel cycle is used, b.
The neutron leakages in the vertical direction and in one of the j
radial directions (X) are neglected while the neutron leakage in I
the other radial direction (Y) is accounted for.
Fixed neutron absorbers in the fuel assembly and soluble poison in c.
the water in case of flooding are neglected.
The calculations to determine the optimum modc ration condition were performed with a Monte Carlo neutron transport code, KENO, with a j
123-group cross section library generated from a basic GAM-THERMOS library using the NITARL routine in the the AMPX code package.
The effects of small changes in rack parameters on rack system criticality I
were determined with a four-group diffusion theory code, PDQ-07, and l
with the neutron cross sections generated from the MUS Corporation version of LEOPARD, NUMICE.
Both PDQ-07 and the KENO calculation methods have been benchmarked.
The criticality analyses of the normal wet condition for the spent fuel racks included several conservative assumptions as well as the effects of uncertainties in the calculation method and geometric and material variations of the fuel storage rack.
The following
'l conservative assumptions were used in the calculation:
1 l
1 28 0661m(041387) 0327A
)
y s
1 Li
.)
Fresh fuel of 4.00'W/o U-235' enrichment - Initially the maximum-
.a.
enrichment will'be lower than this, but'it could approach this enrichment _if.an 18-month fuel cycle is'used.
The. enrichment i) selected is higher than the' average enrichment of any3 fuel; j
expected to be stored in the. spent fuel pool for conservatism..
I
'b.
Spenti fuel pool bulk water temperature - A calculation was' done J o
~
t j
determine the increase in reactivity due to a further decrease in
'j the design-basis temperature'of 68'F'and this reactivity was.added to the nominal.keft for still further conservatism, c.
The neutron leakages in all three directions were negl'cted.-
e l
d.
Fixed neutron absorbers.in the fuel assembly and-soluble poison in
,I the pool water:are neglected.
j The majority.of the calculations were performed with' methods commonly used in light water reactor design, i.e., four group diffusion theory with all calculations using.PDQ-07. Cross' sections for these calculations are generated with NUMICE, the NUS Corporation version of the Westinghouse. LEOPARD code. -This code.uses the same. cross section library tape and calculation techniques as. LEOPARD.
Selected cases were checked and the final design multiplication factors were verified with Monte Carlo' calculations using KENO with a 123-group. cross section library generated from a basic GAM-THERMOS library using the NI1 AWL routine'in the AMPX code package. Both the PDQ-07: and the ICNO calculation methods as described'above have been benchmarked ~.
The criticality analyses for dry storage of new fuel.in spent fueli racks were performed with the assumptions'and: calculational methods.
similar to those utilized for the analyses of new fuel racks. The only exception was that the spent fuel storage rack was assumed to be infinite in size in all three dimensions for additional' conservatism.
2.2.9 Fuel assemblies may be lifted. vertically without any horizontal motion, for inspection purposes, if they.had been stored'in the new i
fuel vault initially without being inspected.'. The new fuel assembly handling tool and a scale for the fuel handling building crane willLbe properly installed and the new fuel assembly handling. tool' coupled to the fuel assembly upper nozzle. After lifting cable slack'is removed, the fuel assembly may be lifted, the protective' cover inspected for damage and removed, and the fuel assembly inspected for damage and
-i lowered back down into its storage rack. These activities will be-performed under direct supervision of a fuel handling' supervisor.
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There will be no more than one fuel assembly outside'its shipping container or storage rack at a given time during receipt and inspection, prior to. initial core loading.
(Two fuel assemblies normally occupy. a shipping container and upon removal _are lif ted together to the vertical position by lifting the strong-back apparatus to the vertical position, but until one is removed from the strong-back both are considered to be located in the shipping container.)
Fuel assemblies may be moved within the Fuel Handling Building in preparation for core loading. During the initial core loading, two fuel assemblies, one in the spent fuel pit area and one in the containment reactor cavity area, will be outside their storage racks at the same time.
2.2.10 Pursuant to 10 CFR 70.24(d), it is requested that Commonwealth Edison Company be exempted from the monitoring requirements of 10 CFR 70.24(a) as they apply to the handling and storage of nuclear fuel assemblies at Braidwood Station.
The procedures and storage 4
facilities described in this application provide assurance that inadvertent criticality cannot occur during receipt, handling, and storage of nuclear fuel assemblies at Braidwood Station.
The Westinghouse model RCC-1 and RCC-3 shippir.g containers are approved for transport of Fissile Class II material as defined in.10 CFR 71.4.
The procedures for unloading and inspecting the fuel are based in part on the prevention of criticality during these operations.
The new fuel storage racks are designed to prevent a critical configuration under conditions of optimum moderation in storage of 4.00 "/o U-235 enriched fuel.
The spent fuel racks are designed to prevent a critical configuration in wet storage of 4.00 "/o U-235 enriched fuel.
Procedural control's will prevent dry storage of'new fuel with greater than 3.20 "/o U-235 in the spent fuel racks and will limit to a checkerboard array any new fuel stored in the racks with enrichment of 3.20 W/o or less.
It is also requested that Commonwealth Edison Company be exempted from the requirements of paragraph 70.24(b) in accordance with paragraph 70.24(c) since Braidwood Station Unit 2 is being built under Construction Permit No. CPPR-133.
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.
Accidental damage to the unieradiated fuel would release relatively -
minor amounts of radioactivity that would be accommodated by the Fuel Handling Building Ventilation System.
Each fuel assembly stored in the new fuel vaults is protected from falling objects by locked steel cover plates, each of which provides protection for up to four fuel assemblies.
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The spent fuel racks were designed to withstand either a dropped fuel assembly or the upward force of a stuck assembly without loss of function.
Two abnormal dropped fuel assembly conditions were analyzed.
They are'(1) a dropped fuel assembly assumed to lay across the-top of a fuel rack, and-(2) a fuel assembly dropped vertically into an empty but unintended (under the checkerboard loading pattern) cell surrounded by four cells already occupied by fuel assemblies (or dropped between adjacent racks immediately adjacent to a cell that is already occupied).
Both cases are not a criticality concern since keft is well below the maximum allowable limit of 0.98.for new fuel storage. All new fuel stored dry in the spent fuel racks will be-protected from dirt and falling objects-by covering the tops of the loaded racks with plastic and two layers of 1/8-inch thick aluminum.
Fuel assemblies will further be protected from falling objects by defining a heavy load exclusion area above stored fuel and by transporting loads along a designated heavy load removal path as described in Section 1.2.3 of this application.
In the event of an accident, the emergency plan of action would be to follow the Commonwealth Edison Company Generating Station Emergency P1.an(GSEP), and take specific recommended actions depending on the severity of the accident.
The design bases and safety evaluation of the new and spent fuel storage facilities is discussed in detail in the Byron /Braidwood FSAR-Sections 9.1.1.1, 9.1.1.3, 9.1.2.1, and 9.1.2.3,-(Attachment G) and in the Byron Station SER (NUREG-0876) and Braidwood Station SER (NUREG-1002), Section 9.1.1,.(Attachment F).
Chapter 3 Other Materials Requiring NRC License Commonwealth Edison Company has already been granted a Facility Operating License, Number NPF-59 to possess various quantities of byproduct, source, and special nuclear matarial for use at Braidwood Station Unit 1.
This license covers fuel assemblies, neutron monitoring fission detectors, calibration sources, and startup neutron sources.
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U.S. NUCLEAR RECULATORY CCTVIMISSION A m tess m3 Q"%.gp# REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH o...
REGULATORY GUIDE 3.15 (Task CE 2194)
STANDARD FORMAT AND CONTENT OF LICENSE APPLICATIONS FOR STORAGE ONLY OF UNIRRADIATED POWER REACTOR FUEL AND ASSOCIATED RADIOACTIVE MATERIAL
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USNRC #EQut.ATOAVGutOCS Comments shouac to sent to the Secretary of the commission.
U. S. Peuclear Regulatory Commissio't. Wasnington, D.C. 20585 Regulatory Gulces are 4 saved to describe anc make evaltante to the Attention: Docketing eno Serv 6ce tranch.
puosec mett.oos acceptap te to the NRC start of implementing specific parts of the Comm6ssion's requestions, to seeineate tech.
eagues used Dy the staff an evaluating specitac proosems or postu.
The pulses are tasued in the followtng ten spread gayesions:
Bated esclaents or to provane puedence to apetecents, mogulatory Guaoes are nol substitutes for reputations, and comettence with
- 1. Power Reactors S. #roducts thIm as not recu4ree. Methods sna sosuttons estforent from snose het
- 2. Research ano Tect Reactors
- 7. Transportation out in the guaces will be accootasse if they prov8ee a bases for the
- 3. Fuels one Materials Facitet6es
- 3. Occupational Health
- 4. Env6tenmental ano 14teng
- 3. Antitrust ang 76nancial Rev6ew 86ndlnt> roovesate to the assuance of continuance of a permet or S. hasterials ans plant protection to. General loconse ey the Commtss#on.
This gulee was ensued after consicerat6on of comments receeved from Cootee of issued guides may be purchased at the current Govemment the puos6c. Comments ano suggestions for amorevements an these printing Off ace orece. A subscription servate for future gutoes en soo-W60es are encourages at all temos, ens guages watt De revise 6. es cific civisions as avallabas through the Government prenten Office.
inverneattin on the subscription service one current GPO or acTroortate. to accommodate comments and to pe sect view enforma.
r e: may tion or eNpertenCo.
De ODtaines Dy wretsne the U.S. Nuclear Regulatory Commus os..
Washington, D.C. 20SSS. Attent6on: Pub 6. cations Sales BAanager.
INTRODUCTION Section 70.3, " License Requirements," of 10. CFR Part 70, " Domestic Licens-ing of Special Nuclear Material," states that no person subject to the regula-tions'.in Part 70 may receive title to own, acquire, deliver, receive, possess, use, or transfer special nuclear material, except as authorized in a license issued by the Commission.
The required content of an application for such license is described in general terms in 5 70.22 of 10 CFR Part'70.
This guide describes the detailed information that is needed by the NRC staff in its review of an application for a license to authorize the receipt, possession, and storage of unirradiated fuel assemblies and associated radioactive materials for eventual use in a nuclear power reactor and suggests a format for.its presentation.
Any guidance in this document related to information collection activities-has been cleared under DMB Clearance No.- 3150-0009.
Purpose and Use This Standard Format and Content has been prepared to encourage a uniformity in these license applications and to provide detailed guidance on the information that should be provided in the applications.
The information provided in the license application must be sufficient to permit a determination to be made of whether the applicant's proposed activities can be conducted without undue risk to the health and safety of the public and the common defense and. security.
The information identified herein represents the minimum that should be provided.
Additional information may be requested if needed for the NRC staff review.
The Standard Format identifies the principal detailed information that the NRC staff needs for its evaluation of the application.. In providing the informa-tion described in the Standard Format, the applicant should use a narrative form of presentation.
The application must provide the information required by 6 70.22 of Part 70.
Applicants may request exemption from the requirements of 5 70.24* as provided-in paragraph 70.24(d).
Proprietary Information Proprietary information should be submitted separately.
When submitted, this information should be clearly identified and accompanied by the applicant's detailed reasons and justifications for requesting its being withheld from public disclosure, as specified by 5 2.790, "Public Inspections, Exemptions, Requests for Withholding,"
of 10 CFR Part 2, " Rules of Practice for Domestic Licensing Proceedings."
Style and Composition The applicant should strive for clear, concise presentation of the in/orma-tica provided in the license application.
" Applicants for a license for a nuclear power plant are exempt from the requirements of paragraph 70.24(b) in accordance with paragraph 70.24(c).
3.15-1 o
References, including author, date, and page number, should be cited within the text if this is important to the meaning of the statement.
Other references used should appear either as footnotes on the page where referenced or at the cnd of each chapter.
Where numerical values are stated, the number of significant figures given should reflect the accuracy or precision to which the number is known. Where appropriate, estimated limits of error or uncertainty should be given.
Abbreviations should be consistent throughout the license application and should be consistent with generally accepted usage.
Any abbreviations, symbols, or special terms not in general usage or unique to the proposed installation should be defined when they first appear in the license application.
NUREG-0544, "A Handbook of Acronyms and Initialisms,"" may be useful.
Graphic presentations such as drawings, maps, diagrams, sketches, and tables should be employed if the information may be presented more adequately l
er conveniently by such means.
Due concern should be taken to ensure that all information so presented is legible, that symbols are defined, and that draw-ings are not reduced to the extent that visual aids are necessary to interpret p:rtinent items of information.
These graphic presentations should be located uith the section in which they are primarily referenced.
The license application should provide the information needed to satisfy the requirements of the NRC regulations, which are codified in Title 10, Chapter I, of the Code of Federal Regulations. As the sections of the application are' developed by the applicant, the applicable regulatory requirements that are being satisfied should be identified.
This procedure will contribute to a more timely review of the presented ira'ormation.
Fnysical Specifications 1.
Paper Size Text pages: 8-1/2 x 11 inches.
Drawings and graphics:
8-1/2 x 11 inches preferred; however, a larger size is acceptable provided the finished copy when folded does not exceed 8-1/2 x 11 inches.
2.
Paper Stock and Ink.
Suitable quality in substance, paper color, and ink density for handling and reproduction by microfilming or image-copying equipment.
3.
Page Marains.
A margin of no less than 1 inch should be maintained on the top, bottom, and binding side of all pages submitted.
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Copies may be obtained at current' prices from the National Technical Informa-tion Service, Springfield, Virginia 22161.
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Printino Composition:
text pages should be single spaced.
Type face and style:
should be suitable for reproduction by microfilming or image-copying equipment.
Reproduction: may be mechanically or photographically reproduced.- All'pages of text should be printed on both sides with image printed head to head.
5.
Binding.
Pages should be punched.for standard 3-hole loose-leaf binders.
Procedures for Updatino or Revising Paces Data and text should be updated or revised by replacing pages.
" Pen and j
ink" or " cut and paste" changes should not be used.
l The changed or revised portion of each page should be highlighted by a
" change indicator" mark consisting of a bold vertical line drawn in the margin opposite the binding margin.
The line should be of the same length as the-portion actually changed.
All pages submitted to update, revise, or add pages to the application-should show the date of change and a change or amendment number in.the lower righthand corner.
A guide.page listing the pages to be inserted and the pages to be removed should accompany the revised pages.
Material Incorporated by Reference j
The application should contain all the information in sufficient detail to permit an independent health and safety evaluation of the proposed activities.
Information contained in previous communications filed with the Commission by the applicant or the fuel supplier, including any information submitted in connection with the construction permit or operating license application, may be incorporated into the application by reference (applicable portions of the Final Safety Analysis Report (FSAR) should be attached),'provided such references are clear-and specific and the applicability of the referenced i
material is demonstrated.
References that are not available in the Public Document Room should be provided with the application. An application that is self-contained with respect to technical data can usually be processed more expeditiously than one requiring many referrals to other documents.
Distribution Eight copies of the application in letter form should be submitted.
The application should be signed by a corporate officer but need not be notarized.
Copies should be filed with the Director of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555._ Sub-sequent correspondence should reference the 10 CFR Part 70 docket number.
Fees There is no license fee for a license issued for storage only of fuel to be used in a reactor for which a construction parait has been issued.
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Chapter 1 GENERAL INFORMATION 1.1 Reactor and Fuel This section should provide information regarding the reactor and the fuel.
1.
Identify the reactor, all parties (utilities) to the licensing action, its geographic location, and the docket and construction permit numbers.
2.
Describe the fuel assemblies, including materials of construction; number of fuel rods; diameter of fuel pellets; cladding thickness and outside diameter; rod pitch; the number and location of instrument, water, or gas channels in the assemblies; and other appurtenances such as control rod guide tubes, spacer rods, and burnable poisons.
3.
State the maximum enrichment (pin values) of uranium; maximum quan-tities (kg) of U-235, U-233, plutonium, natural uranium, depleted uranium, and thorium per assembly; and the total weight of the assembly.
j 4.
State the total number of fuel assemblies for which a license is roquested and the total weight of U-235, U-233, plutonium, natural uranium, depleted uranium, and thorium contained therein.
l 1.2 Storage Conditions This section should provide information regarding the storage conditions.
1.
Provide scale drawings showing the areas where fuel assemblies will be stored and, if appropriate, inspected and channeled.
2.
Describe the storage environment.
3.
Describe the nature of the activities conducted in all adjacent ar,eas and the potential effects of these activities on the safety of storage.
4.
Describe the storage facility structures, components, equipment, and systems (e.g., racks, cranes, inspection stands) and provide the design crite-ria used to ensure structural integrity.
5.
Describe the fire alarm and fire control systems.
6.
Describe the controls for preventing unauthorized access to areas where special nuclear material is stored.
1.3 Physical Protection If the quantity of U-235, U-233, or plutonium to be possessed under the license is equal to or greater than the quantity specified in paragraph 73.1(b) i 3.15-4
b of 10 CFR Part 73, " Physical Protection of. Plants and Materials," the licensee must comply with the requirements of that regulation.
Specific guidance for applicants regarding physical protection will be provided by the Commission.
1.4 Transfer of.Special Nuclear Material This section should provide the required information regarding the transfer, control, and accounting of special' nuclear material; 1.
If the fuel fabricator or other organization. is responsibis for the shipment ~of fuel to the applicant, identify the responsible. shipper.
2.
If the applicant is responsible for the packaging of fuel for delivery to a carrier for. transport,. information concerning Lthe packaging and '
transfer.should be provided in accordance with.the provisions of 10 CFR Part 71, " Packaging of Radioactive Material for. Transport and Transportation of Radioactive Material Under Certain Conditions."
3.
Describe the special' nuclear material control and accounting prac-tices to be implemented to comply with the applicable provisions of 10 CFR-Part 70.
1.5 Financial protection and Indemnity Requirements for financial protection to be provided by licensees are set-forth in 10 CFR Part 140, " Financial Protection Requirements.and Indemnity
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Agreements." This section should provide the required information regarding financial protection.
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Persons subject to Subpart B (private organizations) should furnish proof of financial protection (see $ 140.15) in the amount required by.
$ 140.13, 2.
Persons subject to Subpart C (Federal agencies as defined in para-graph 140.3(c)) are not required to furnish financial protection (see $$ 140.51 and 140.52).
3.
Persons subject to Subpart 0 (nonprofit educational institutions) are not required to furnish financial protection (see~55 140.71 and 140.72).
The applicant'should submit a statement certifying that the applicant is a nonprofit educational institution and that the license for which application is made will be used in connection with,the conduct of educational. activities.
Note: With respect to each category cf licensee in Section 1.5, the Commission i
E execute and issue agreements of indemnity pursuant to the regulations in 10 CFR Part 140.
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Chapter 2 HEALTH AND SAFETY j
i 2.1 Radiation control This section should provide information regarding the radiation control
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program.
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1.
State the minimum qualifications for the key positions having radiation safety responsibilities.
2.
Describe the responsibilities for the key radiation safety personnel.
3.
Indicate the training and experience of the person,or persons respon-sible for radiation safety.
4.
Describe the procedures and equipment for checking for contesination and the steps that will be taken if contamination is detected.
5.
State the frequency and methods for calibrating and testing radio-Icgical protection instruments.
6.
Describe the procedures and equipment to be used to meet applicable ssctions of 10 CFR Part 20 " Standards for Protection Against Radiation."
7.
Describe the provisions for disposal of any radioactive wastes that cre generated.
f 2.2 Nuclear Criticality Safety This section should provide the required information regarding nuclear criticality safety.
1.
State the minimum qualifications for the key positions having nuclear criticality safety and fuel handling responsibilities.
2.
Describe the responsibilities for the key personnel responsible for nuclear criticality safety and fuel handling.
3.
If fuel elements will be stored in shipping containers, describe the containers, the storage array, and the basis for nuclear criticality safety.of the container storage array.
4.
If elements will not be stored in their shipping containers, provide a nuclear safety analysis of the alternative storage, including a description of the physical means for maintaining safe spacing of the elements (e.g., stor-age racks) and controls to be exercised over placing the elements in the stor-cge location.
In situations in which nuclear safety depends on proper spacing, j
the integrity of storage fixtures should be analyzed from the standpoint of possible failure due to such factors as loading, shock, fire, or corrosion.
Provide drawings or sketches of racks and provide the spacing between elements and between elements and floors or walls.
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5.
If nuclear criticality safety is based on other than the maximum-
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enrichment of the fuel, identify and justify the nominal enrichment used in the i
safety analysis.
6.
If nuclear criticality safety is based on the reactivity effects of i
neutron absorber materials in the racks, describe the chemical and physical' properties of the materials.
Drawings or sketches showing the dimensions and i
locations of the neutron absorber materials should be provided.
Describe quality assurance activities to ensure the continued presence and effectiveness j
of the neutron absorber materials.
7.
If nuclear safety is based on moderation control, sources of. water that go into dry storage.(e.g., sprinkler systems) and the. probability'of 4
flooding the storage area should be taken into account in the analysis.
The nuclear safety analysis should demonstrate that fuel will be stored in such a manner that if the fuel were flooded and then drained, water could not be retained around or within an assembly; however, if water retention is possible, i
its results should be evaluated.
The analysis should also evaluate the effects i
of low-density moderators (e.g., mist) or show that such densities are not credible.
8.
Describe the validation of the calculational method for nuclear crit-icality safety.
For guidance on validation of calculational methods, see Regu-latory Guides 3.4, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," and 3.41, " Validation of Calculational Methods for Nuclear Criticality Safety."
9.
If elements will be removed from storage (e.g., for inspection pur--
poses), describe in detail the activities to be performed and the controls to be exercised over removing and replacing the elements.
The maximum number of fuel assemblies that will be out of approved shipping containers or approved
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storage racks at any one time should be stated; this number should be justified, i
preferably on the basis that it is less than the minimum number required to l
achieve criticality under optimum conditions of spacing, moderation, and reflection.
As an alternative, the applicant may justify the safety of a large number of assemblies on the basis of a nuclear safety analysis that includes the considerations in paragraphs 4 and 7 of this section.
10.
If an exemption from the requirements of f 70.24 is not requested or has been disapproved, describe the plans for compliance with the requirements of 6 70.24,* including instrumentation, location of detectors, and. emergency procedures and drills.
2.3 Accident Analysis This section should identify and evaluate potential accidents that could affect the safety of storage, such as dropping fuel. assemblies or other objects over the storage area or dropping fuel assemblies at other places, and the i
emergency plan of action if such events should occur.
m Applicants for a license for a nuclear power plant are exempt from the require-ments of paragraph 70.24(b) in accordance with paragraph 70.24(c).
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i Chapter 3 OTHER MATERIALS REQUIRING NRC LICENSE If source or special nuclear materials other than those contained in fuel i
assemblies or if byproduct materials are to be possessed at the storage site (e.g., fission chambers, calibration sources, or startup sources), the follow-ing information should be provided-1.
Identify the type and amount of material and conditions of storage.
2.
Describe the use if other than storage.
3.
Describe radiation protection provisions.
4.
Describe the control and accounting procedures for other special nuclear material.
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VALUE/ IMPACT STATEMENT The original version of Regulatory Guide 3.15 was issued in October 1973.
Many of the applications by utilities for " Storage Only" licenses are incomplete, particularly rega-ding the information required to perform an independent nuclear criticality safety analysis of the fuel storage arrangements.
The revised guide should assist the applicant in preparing a more complete applica-tion and reduce the amount of additional information the NRC must request of the applicant.
The revised guide also updates the physical protection requirements.
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B/B AMENDMENT 7 APRIL 1986 2.3.12 Fuel Handlina Building 2.3.12.1 Fuel Handlino Buildina ' Fire Zone 12.1-Ci This zone is shown on Figures 2.3-19 and 2.3-20 and in elevation on Figure 2.3-22 ' Sheet 2).
j Fire Barrier Description All slabs and exterior walls in the Fuel Handling Building are structural reinforced concrete.
The floor slab at elevation 401 feet 0 inch varies from 55 inches to 96 inches thick structural reinforced concrete and is the foundation.
The floor slab at elevation 411 feet 0 inch, which is the fuel unloading platform, is 18-inch structural reinforced concrete and is between column 4
17 and 19.6.
The platform is supported by concrete columns.
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This platform extends to row Y, but is covered by the new fuel storage area at elevation 426 feet 0 inch from row Y to 2.
The north, south, and west exterior walls are 36-inch thick structural reinforced concrete.
The east wall, which includes the Spent Fuel Storage Pool wall up to elevation 426 feet C inch, is structural reinforced concrete or solid concrete mansonry units and varies in thickness from 12 inches to 84 inches.
It separates the Fuel Handling Building from the Auxiliary Building and carries a 3-hour fire rating.
It contains three Label "A" fire doors, and two pairs of doors which are either of Label "A" construction which close tightly around a continuous steel trolley beam (Byron), or are Label "A" fire doors (Braidwood).
A portion of the auxiliary building floor slabs at elevation 451 feet 0 inch and elevation 426 feet 0 inch also separates the Fuel Handling Building from the Auxiliary Building.
These slabs are a minimum 15-inch clear cover of structural reinforced concrete over 3-inch fluted steel decking formwork, supported by steel beams protected by a fire resistant covering and columns, and carry a 3-hour fire rating.
All exterior walls extend up to and are dowelled into the roof slab.
The roof slab at elevation 474 feet 6 inches is 14-inch over 3-inch metal decki'ng which is supported by unprotected structural steel beams and columns.
The Spent Fuel Storage Pool extends from elevation 385 feet 0 inch to elevation 426 feet 0 inch and is open to the operating floor.
The floor slab and walls of the pool are structural reinforced concrete.
The thickness of the walls varies from 60 inches to 84 inches, and the floor slab is 65 inches thick with a 7-inch concrete topping.
A continuous stainless steel liner plate is anchored to the floor slab and walls of the pool.
The floor level at elevation 401 feet 0 inch includes the following areas:
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a.
Heat Exchancer Area The walls-are minimum 20-inch thick solid concrete masonry units, structural reinforced' concrete, or removable solid concrete masonry units.
The floor slab is an 8-inch concrete topping over a 55-inch thick structural reinforced concrete'which is the grade foundation...Two.36-inch by.36-inch' drainage sumps are located ~in the northwest and southwest corners of the heat exchanger, areas.
b.
Soent Fuel Pump Room The walls are minimum 20-inch thick solid concrete The base slab ~at elevation:401 feet 4 masonry units.
inches is an 8-inch concrete topping over.64-inch thick structural reinforced concrete, c.
Decontamination Area i
The decontamination area extends'from elevation 401 feet 0 inch up to elevation 426-feet 0 inch and is open to the operating floor.
The floor slab at elevation 401 feet 0 inch is 62-inch structural reinforced concrete.
The walls.are 24-inch thick structural reinforced concrete.
A continuous stainless steel plate is anchored to the walls and floor slab of the decontamination pit.
A 36-inch by 36-inch drainage sump is located _in the' northeast corner of this area Sixteen 4-inch floor drains (90 gpm capacity each) are provided in the heat exchanger area, spent fue1< pump room, fuel unloading area, and the railroad track area.
These drains flow.to the heat exchanger area sump and are then routed to the Radwaste Building.
The intermediate floor level at elevation 411 feet 0 inch, which includes the new fuel unloading area and new fuel storage area,.
l is an 18-inch structural reinforced concrete slab and also is the l
This slab roof of the heat exchanger and spent fuel pump areas.
is supported by concrete columns.
A 4-inch concrete curb is provided at the edge of the slab as well as removable handrails.
Two 4-inch diameter drains are provided in the new fuel unloading and four drains (90 gpm each) in the new fuel storage area.
- area, These drains flow to the heat exchanger area sump.
The new fuel storage area is enclosed by minimum.24-inch thick
.i structural reinforced concrete walls which extend up to and are dowelled into the floor slab at elevation 426 feet 0 inch.
The operating floor level at elevation 426 feet 0 inch is an 18-inch thick structural reinforced concrete slab.
There'are a l
number of openings through this floor slab.
The openings.to the spent fuel transfer canal and the new fuel storage area are 2.3-153 1
1 B/B AMENDMENT 7 APRIL 1986 a
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covered by removable checkered plates.
The other. openings are to equipment cubicles and are covered by removable concrete slabs.
The railroad car shed is attached to the Fuel Handling Building.
j The floor slab of the car shed area at elevation 401' feet 0 inch 1s 36-inch thick structural reinforced concrete and the walls are insulated metal siding.
The roof slab at elevation 426 feet 0 inch is precast concrete roof slab supported by unprotected structural steel beams and columns.
Two nonrated rolling steel doors are provided.in the car shed and lead to the Fuel Handling Building and to the outside respectively.
The railroad tracks run through the car shed and into the Fuel Handling Building.
Duct penetrations in the floors do not have fire dampers installed.
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It has been determined this lack of dampers will not prevent safe shutdown and is deemed satisfactory for property loss reascns.
Fire dampers are provided in the fire wall separating the fuel i
handling building and the auxiliary building.
Piping penetrations are sealed.
Safety-Related Equipment Safety-related equipment located in the Fuel Handling Building consists of the components of the Auxiliary Building Ventilation Svs. tem, the 125-ton fuel handling crane, spent fuel pit heat excnanger, and the fuel pit sump pump.
Only the spent. fuel pit heat exchanger and pumps are needed to maintain fuel pool temperature.
The cranes, spent fuel pit heat exchangers, and pumps are designed as Seismic Category I.
The HVAC safety-related equipment located in this fire area are components of Auxiliary Building Ventilation System.
These include galvanized ductwork, ductwork accessories, and cubicle coolers.
Protection Criteria and Measures The protection measures for this area consist of separation of the area from the rest of the plant by a fire barrier.
The new l
fuel storage racks are designed to preclude criticality problems when subjected to water spray.
Nonfogging type nozzles are used for the manual hose station located in this area to preclude conditions of' optimum moderation for the new fuel storage racks.
If a rupture should occur, no safety-related equipment is likely to be damaged.
Fire dampers are provided in the fire wall.
Fire dampers are fusible link type which will be closed when the temperature exceeds 165' F.
2.3-154 1
AMENDMENT 7 B/B APRIL 1986 Combustible Materials The building crane, 125 ton capacity, contains an estimated 5 pounds of grease and 35 gallons of lubricating oil and the spent fuel bridge crane about 41 pounds of grease and 3 pints of lube oil.
Miscellaneous pumps contain less than 1 gallon of oil.
Each row of new fuel storage racks in the new fuel storage area on elevation 411 feet 0 inches has a 2-inch thick by 10-inch wide This floor of Douglas Fir on which the new fuel assemblies rest.
wood floor is estimated conservatively to contain 40 cubic feet of wood.
At a density of 34 pounds per cubic foot, and a hegt content of 8000 Btu /lb, the resulting heat load is 10. 9 x 10 Btu.
Also present during construction of Unit 2 is a wooden railroad structure on elevation 426 feet 0 inches.
This railroad support is used for transporting materials into and out of the containment from the area next to the spent fuel pool on 426 feet 0 inches.
This material is not listed in Table 2.2-3.
It is considered to be included in the transient fire load.
No special fire protection features are considered necessary.
A transient load is considered for this zone.
Materials such as wood shipping crates for equipment and fuel storage racks are occasionally stored on the elevation 401 feet 0 inch floor.
Trucks shipping new or spent fuel would occasionally be present.
The approximate amount of this combustible material present is estimated to be equivalent to four 55-gallon drums of oil.
There is 1 pound of lubricating grease associated with each of the cubicle cooler motors.
The ductwork contains 213 pounds of gasketing and 59 pounds of insulation.
Fire Loadino The majority of the combustible materials are located in an 87 foot by 72 foot area bounded by column rows Y and BB and 15 and 21 at elevation 401 feet 0 inch.
With a total floor area of 10,680 ft2, the fire loading is about 5.,000 Btu /f t2 Extinouishina and Detection Capability Portable fire extinguishers and four manual hose stations are located at elevation 426 feet 0 inch, Figure 2.3-19 and Figure 2.3-10, sheets 1 and 3, and two manual hose stations and portable fire extinguishers are provided at elevation 401 feet 0 inch, Figure 2.3-20.
2.3-155 l
AMENDMENT 6 B/B JANUARY 1985 Similar equipment can be brought in f rom the Auxiliary Building at elevation 426 feet 0 inch if needed.
Ionization detectors are provided for elevation 401 feet 0 inch, 411' feet 0 inch and 426 feet 0 inch which alarm and annunciate in the Control Room.
The large area above elevation 426 feet 0 inch is.provided with ultraviolet detectors which alarm and annunciate in the Control Room.
Desion-Basis Fire Due to the-quantity of combustibles present, the design-basis fire would be localized to the floor area at elevation 401 feet 0 inch and of short duration The spread of heat and combustion products would be mitigated by the fire barriers.
The available manual extinguishing equipment is capable of rapidly extinguishing the design-basis fire.
The only equipment that would conceivably be damaged by the l
design-basis fire would be the safety-related Fuel Handling Building Ventilation System.
These systems and components are l
not required for safe shutdown and cooldown of the reactor.
If the temperature exceeds 165' F, the fusible links will melt and the fire dampers will close.
2.3-156
BRAIDWOOD AMENDMENT 7
'cRIL 1986 IMPLEMENTATION OR JUSTIFICATION FOR SECTION NRC POSITION NONCOMPLIANCE g.
Govern leak testings siellar procedures Comply. Leak testing is done with cuch as airflow determination should noncombustible materials.
use one of the commercially available techniques. Open flames or combustion-generated smoke should not be permitted.
h.
Maintain the periodic housekeeping Comply. Refer to Section III.K.7 inspections to ensure continued com-of Appendix AS.7.
pliance with these administrative controls.
- i. Control the use of specific combustibles Comply with exceptions. Refer to in safety-related areas. All wood Section III.K.8 of Appendix AS.7 for used.in sr.fety-related areas during
- details, maintenance, modification, or refueling cperation (such as lay-down blocks or scaffolding) should be treated with a flame retardant. Equipment or supplies (such as new fuel) shipped in untreated combustible packing containers cay be unpacked in safety-related creas if required for valid operating reasons. However, all combustible materials should be removed from the area immediately following unpacking.
Such transient combustible material, unless stored in approved containers, should not be left unattended during lunch breaks, shift changes, or other similar periods. Loose combustible packing materials cuch as wood or paper excelsior, or polyethylene sheeting should be placed in metal containers with tight-fitting self-closing metal covers.
- j. Disarming of fire detection or fire Comply. Disarming of tire detection suppression systems should be con-or fire suppression systems is controlled crolled by a permit system. Fire by BWAP 1100-8,
- Fire Protection Impair-watches should be established in areas ment Procedure," which notes that l
where systems are so disarmed.
the Shif t Engineer shall ensure that the fire protection area that is re-moved from service is protected by extra fire equipment and that fire watches are established in areas where detection or suppression is impaired.
A permit system is utilized to control impairments, k.
Succet.sf ul fire protection requires C omply. The scope of the job classifi-testiaq and maintenance of the fire cation for maintenance personnel identi-protecraon equipment and the emergency fies job responsibilities in the area lighting and communication. A test of fire protection.
plant taat lists the individuals and their responsibilities in connection BWAP 1100-8 " Fire Protecticn Impairment,"
vi th routine tests and inspections provides instructions in tie use of oL the fire detection and protection fire protection impairment permit cards systems should be developed. The in the event a fire protection component test plan should contain the types, is taken out of service. The surveil-frequency, and detailed procedures lance program (Test Plan) contains for testing. Procedures should also the types, frequency, and detailed contain instructions on maintaining procedures for testing.
fire protection during those periods I
when the fire protection system is impaired or during periods of plant maintenance, e.g.,
fire watches or temporary hose connections to water systems.
3.2-2
BRAIDWOO3 AMENDMENT 7 APRIL 1906 IMPLEMEWyATION OR JUSTIFICAT!DW FOR SECTION NRC P001710N NONCOMPLIANCE 3
Diesel Fuel 011 Storace Aress Diesel fuel oil tanks with a capac1ty Comply, escept es noted belows greater than 1,100 gallons should not be located inside buildings Diesel fuel oil tanks are located containing safety-related equipment.
' within the musillary building. See If above-ground tanks are used, they Table 3-1 for delineation of confor-should be located at Isast 50 feet
. mance with NFPA 30.
from any building containing safety-related equipment or, if located
'j within 50 feet, they should be housed in a separate building with construc~
4 tion having a minimum fire resistance j
rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Potential oil spills should be confined or directed 4
away f rom buildings containing saf ety-related equipment. Totally buried J
tanks are acceptable outside or under buildings (see NFPA 30, ' Flammable and Combustible Liquids Code,' for additional guidance).
Above-ground tanks should be protected by an automatic fire suppression i
system.
I k.
Safety-Related Pumps Pump houses and rooms housing redun-Comply, escept as noted below:
1 dant saf ety-related pump trains should be separated from each other and Most saf ety-related pumps are located f rom other areas of the plant by in the sus 111ary building. Most pumps fire barriers having at least 3-hour are located in individual rooms separated ratings. These rooms should be pro-from other plant areas by walls of i
tected by automatic fire detection substantial construction, but which I
and suppression unless a fire hasards generally do not carry fire ratinos, j
analysis can demonstrate that a fire Automatic fire detection is provided a
will not endanger other safety-related for all pumps, but automatic suppression equipment required for safe plant is in general not provided. Refer i
shutdown. Fire detection should to the applicable portions of Sections alarm and annunciate in the control 2.3 and 2.4 of this fire Protection room and alarm locally. Rose stations Report for a description of individual and portable extinguishers should pumps and the fire hasards and safe be readily accessible.
shutdown analyses. Deviations f rom the 10CFR50 Apper.dia R criteria are listed in Table AS.1-1.
Ionisation detectors are provided which annunciete and alarm in the control r oom.
Detector $ do not alarm locally.
j Floor drains should be provided to prevent water accumulation f rom dam-aging safety-related equipment (see Pos i tion C.S.a. (14) ).
Provisions should be made for manual control of the ventilation system to facilitate smoke removal if requised for manual firefighting operation i
(see Position C.5.fl.
1.
New Fuel Af ea Rand portable extinguishers should Comply, except as noted belows be located within this area. Also, hose stations should be located outside Ionisation and ultraviolet detectors but within hose reach of this area.
are provided which annunciate and alarm Automatic fire detection should alarm in the control room.
and annunciate in the control room and alarm locally. Combustibles should Detectors do not alarm locally.
be limited to a minimum in the new fuel area. The storage area should be provided with a drainage system to preclude accumulation of water.
3.7-10
APRIL 1986 IMPLEMENTATION OR JUSTIFZCAT2ON FOR SECTION NRC POSITION NONCOMPLIANCE The storage configuration of new fuel should always be so maintained as to preclude criticality for any water density that might occur during fire water application.
m.
Spent Fuel Pool Area Protection for the spent fuel pool Comply, except as noted below area should be provided by local hose stations and portable extinguishers.
Ionization and ultraviolet detectors Automatic fire detection should be are provided which annunciate and alarm provided to alarm and annunciate in the control room.
in the control room and to alarm locally.
Detectors do not alara locally.
n.
Radweste and Decontamination Areas Fire barriers, automatic fire suppres-Comply, except as noted below:
l sion and detection, and ventilation l
controls should be provided.
Radweste areas within the auxiliary building are separated from other auxiliary building arens by nonfire-rated walls. The radwast.e areas within the service building are provided with
- 3-hour rated barriers between the adja-i cent turbine building and between other service building areas. Most radwaste areas are not prov!.ded vith automatic fire suppression systems. Automatic suppression with c9ly partial coverage is provided for radwaste areas within the service building. Refer to Subsec-tion 2.3.14 for detailed description of the various radweste areas of the plant.
Decontamination areas are not treated as separate fire areas, and hence do not have fire betriers or automatic suppression systems. Four such areas are identified on plant drawings. A decontamination skid is shown on the ground floor of the service building.
It is part cf fire zone 14.6-0.
A decontamination station is shown in the radwsste-tunnel on Elevation 383 feet 0 inch near column-row 26/R.
It is part of fire zone 14.1-0.
In the fuel handling building, a decontamination area is shown on Elevation 401 feet 0 inch near column-row 17/Z. It is a pit for washing down spent fuel casks and has neither rated fire barriers nor automatic suppression. It is part of fire zone 12.1-0 (refer to Subsection 2.3.12).
Finally, a decontamination /
change area and a decontamination pad are shown on Elevation 426 feet 0 inch in the general area of the auxiliary building between column-rows 15-21/U-V.
These rooms are separated frcm the rest of the auxiliary building by nonfire-rated walls, and they do not have automatic suppression. Refer to Subsec-tion 2.3.11 for a detailed de. set iption of these areas.
3.7-11
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ATTACHMENT C
l I