ML20134A952
| ML20134A952 | |
| Person / Time | |
|---|---|
| Site: | 07003011 |
| Issue date: | 07/31/1985 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20134A939 | List: |
| References | |
| NUDOCS 8508150476 | |
| Download: ML20134A952 (35) | |
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HOPE CREEK GENERATING STATION Application for License For Storage Only of Unirradiated Nuclear Fuel Revision 2 Public Service Electric and Gas Company, pursuant to j
Title 10, Code of Federal Regulations, Part 70, hereby applies for a license to permit the receipt, possession, 1
inspection, and storage of special nuclear materials in the form of unirradiated nuclear fuel.
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CHAPTER 1 -
GENERAL INFORMATION 1.1 REACTOR AND FUEL 1.1.1 Identification of Reactor, Parties to the i
Licensing Action, Geographic Location, Docket i
and Construction Permit Numbers The application for Special Nuclear Material (SNM) License is submitted by the Public Service Electric and Gas Company (PSE&G) for the nuclear power facility designated " Hope Creek Generating Station" (HCGS).
This station is a one-unit nuclear power plant utilizing a General Electric (GE) Mark I Containment and BWR 4/5 Nuclear Steam Supply System (NSSS) with rated core thermal power of 3293 MWt 9100% steam flow, with gross electrical output of approximately 1118 MWe and net electrical output of approximately 1067 MWe.
Additional general information pertaining to HCGS is located in Chapter 1 of the HCGS FSAR, with specific information q
reference provided therein.
PSE&G is incorporated in the State of New i
Jersey with its' principal of fice located at 80 Park Plaza, Newark, New Jersey 07101.
All directors and principal officers names 1
are listed in Table 1.1.1-1 All are American citizens.
To the best knowledge of the applicant PSE&G is not owned or controlled by any alien, foreign corporation or foreign government.
PSE&G owns an undi-vided 95% interest in the HCGS.
Atlantic City Electric Company owns an undivided 5%
interest in the facility.
Atlantic City Electric Company is incorporated in the State of New Jersey with principal of fices located at 1199 Blackhorse Pike, Pleasantville, New Jersey 08232.
All directors and principal officers for the Atlantic City Electric Company are identified i
in Table 1.1.1-2, all are American Citizens.
i To the best knowledge of the applicant Atlantic City Electric Company is not owned or controlled by any alien, foreign corpora-tion or foreign government.
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2 The HCGS is located on the southern part of Artificial Island on the east bank of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey.
While called Artificial Island, the site is actually connected to the mainland of New Jersey by a strip of tideland formed by hydraulic fill from dredging operations on the Delaware River by the U.S. Army Corps of Engineers.
The site is 15 miles south of the Delaware Memorial Bridge,18 miles south of Wilmington, Delaware, 30 miles southwest of Philadelphia, Pennsylvania, and 7-1/2 miles southwest of Salem, New Jersey.
On November 4, 1974, the Atomic Energy Com-mission, predecessor to the Nuclear Regula-j tory Commission ("NRC") issued Construction Permit No. CPPR-120 to PSE&G and the Atlantic City Electric Company in Docket Number 50-354 for the Hope Creek Generating Station.
1.1.2 Puel Assemblies Each fuel assembly consists of a square chan-nel enclosing an 8 x 8 array (bundle) of Zircaloy rods.
Each bundle consists of sixty-two fuel rods and two water rods for a total of sixty-four rods per bundle.
The rods are supported by an upper and lower tie plate cast from Type 304 stainless steel.
Eight of the fuel rods on the bundle peri-phery are tie rods.
Both threaded ends of these rods pass through the tie plates and are bolted to support and maintain bundle geometry.
Finger springs are located between the lower tie plate and the channel for con-trolling bypass flow.
Each bundle contains two centrally located water rods, one of which is a spacer capture rod designed to provide axial support for seven Zircaloy-4 fuel rod spacers.
The fuel spacers are f abricated from Zircaloy-4 with Inconsi-X i
spring.
The fuel rod spacers laterally sup-port the bundle rods, maintain rod spacing and geometry, as well as dampen any flow j
induced vibrations.
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3 The fuel channels are fabricated from Zircaloy-4.
The channels prevent cross-flow between bundles, guide and provide a bearing surface for control rods, and provide rigid lateral support for the fuel bundles.
The channel is open at the bottom and makes a sliding seal fit on the lower tie plate surface.
At the top of the channel, two diagonally opposite corners have welded tabs, one of which serves as the attachment point to a raised post on the upper tie plate.
The post has a threaded hole to which is attached a channel fastener assembly.
Other pertinent data as required by Regula-I tory Guide 3.15 " Standard Format and Content of License Applications for Storage Only of Unirradiated Power Reactor Fuel and Associ-ated Radioactive Material" is identified in Table 1.1.2-1.
Additional descriptive information pertaining to fuel assembly design, including materials of construction can be located in HCGS FSAR Section 4.2 " Fuel System Design" and refer-ences noted therein.
1.1. 3 Enrichment There are five bundle types in the initial core loading of HCGS.
The number of bundles and nominal concentrations are presented in Table 1.1.3-1.
The fuel bundles contain no U-233, plutonium, depleted uranium, or thorium.
The total weight of a fuel assembly is approximately 700 pounds.
The weight of a fuel bundle is approximately 600 pounds.
1.1.4 Total Nuclear Fuel Material i
Based upon the data presented in Table 1.1.3-1, a total of 764 assemblies containing
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approximately 2582 kg of U-235 will be received.
As such, Public Service Electric and Gas Company hereby requests a license for M P84 153/01 5-az
4 3000 kg of U-235 to allow for manufacturing tolerances and for the receipt of spare i
assemblies if required.
i No licensing request is submitted for U-233, plutonium, depleted uranium or thorium.
1.2 STORAGE CONDITIONS 1.2.1 Storace Locations The principal location where fuel bundles or fuel assemblies will be stored is the Spent Fuel Pool located in the Reactor Building.
Appropriate descriptions and drawings of this area are provided in Sections 1.2 and 9.1.2 of the HCGS FSAR.
Circumstances may arise which could interrupt off-loading and receipt of fuel in the Rail-road Access area.
For example, maintenance of the polar crane or construction activities which conflict with fuel receipt could disrupt fuel receiving activities.
In the event such circumstances occur, fuel may be temporarily stored in the Railroad Access Area on elevation 102' of the Reactor Building.
If equipment malfunctions or other delays occur, one loaded trailer could be parked in the Railroad Access Area.
In addition, there exists the possibility of unloading other containers into this area concurrent with the loaded trailer.
Appropriate drawings and description of this area are provided in Sections 1.2 and 9.1.4.2.10 of the HCGS FSAR.
1.2.2 Storace Environment l
The fuel will be stored channeled or i
unchanneled in the spent fuel racks, dry.
If necessary, the fuel can be stored wet within the spent fuel racks in the spent fuel pool.
1.2.3 Adjacent Area Activities No operation's other than fuel and component inspection, handling, and storage will be l
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performed in the fuel storage area on elevation 201'.
Crane operations will be restricted such that no more than one channeled fuel assembly or equivalent weight per crane will be allowed over storage areas containing f uel.
Loaded fuel shipping con-
.,g' tainers or properly designed overload test weights may be handled in these areas pro-vided that they are at no time suspended over the f uel arrays in storage.
Any non-f uel related activities which must be conducted in the f uel handling area will be reviewed and i
approved by the Shif t Supervisor, or his designee.
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Any non-f uel activities in the Railroad i
Access area on elevation 102'-0" of the Reac-tor Building will be restricted as follows l
during fuel handling:
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a.
No painting, grinding, sandblasting, or similar activities are allowed.
l b.
No overhead work is allowed.
c.
No crane operations other than those required for fuel handling and inspection are allowed.
d.
No construction or test activities which may adversely af fect fire protection in the fuel handling area are allowed.
When fuel handling activities are not in progress, selected activities such as those above may be performed provided
- the fuel is protected and the activities are reviewed and approved by the Shif t Supervisor or Mainte-nance Supervisor or their designee.
Activities in other areas of the Reactor Building need not be restricted during any,of these periods.
1.2.4 Description of storace Facilities
- 1. 2. 4.1 Spent Fuel Pool - Reactor Building The spent fuel pool in the Reactor Building contains a storage space
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. sufficient for 4084 fuel assemblies.
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6 The spent fuel storage racks are constructed in accordance with Seismic Category I requirements.
The applicable code for the design of racks is ASME Section III, Subsection NF.
The spent fuel racks are constructed of ASTM A-240 and ASTM A-564 stainless steel.
The A-240 and A-564 material specifications are identical to the ASME SA-240 and SA-564 material specifications.
All rack steel is supplied with certified material test reports.
The spent fuel storage racks use a neutron absorber to maintain suberiticality in a high density array.
A sufficient quantity of racks will be installed prior to receipt of new fuel on site so that the initial core load can be stored.
The racks are designed to protect the fuel assemblies from excessive physical damage under normal or abnormal conditions.
The racks are constructed in i
accordance with the QA requirements of 10 CFR 50, Appendix B.
I Brooks & Per' kins Corp. manufactures the Boral utilized'by PSE&G in the Hope Creek Generating Station Spent l
Fuel Racks.
Assurance that the Boral composition meets design
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specifications is achieved through use of B&P's procedure BP-100530AP.
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6a Material traceability is maintained i
for the aluminum skins, the boron I
carbide and the aluminum powder from the raw material stage through
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manufacturing.
Work order numbers, weight of B4C and aluminum used, control numbers, and batch numbers are recorded on data sheets that are part of this procedure.
Samples of each batch are lab analyzed to prove homogeneity of the Boral matrix.
After manufac-ture of the sheets, the sheet serial number and the B4C lot numbers are inputted into a com-puter.
This computer data base assures traceability of the mate-rials used at Hope Creek throughout its life cycle.
Boral is a clad composite of boron carbide (B4C) and 1100 alloy alum-inum.
The boral panel consists of three distinct layers.
The outer protective layers are solid alum-in ua.
The central layer contains a uniform aggregate of fine boron carbide particles tightly held within an aluminum alloy matrix.
The boron carbide particles in the central layer average 85 microns in diameter with an average spacial separation of 1.25 to 1.50 particle diameters.
The chemical composition of alumi-num (1100 alloy) is as shown in Table 1.
The chemical composition of Boron Carbide is as shown in Table 2.
The minimum weight per-cent B4C is shown in Table 2 as 94.0.
The ~ minimum Boral density is 2.51 gm/cc - 0.0907 lb/cu. in.
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1 6b Table 1 Chemical Composition - Aluminum (1100 Alloy) 1 99.00% min. - Aluminum 1.00 % max. - Silicon and Iron
.05.20% max. - Copper
.05% max. = Manganese
.10 % max. - Z inc
.15% max. - others each Table 2 Boron Carbide Chemical Composition, Weight %
l Total Boron 70.0 min.
I B10 isotopic content in natural boron 18.0 min.
Boric oxide 3.0 max.
Iron 2.0 max.
Total boron plus-total carbon 94.0 min.
3 Programmed and Remote Systems /GCA Corporation (par) designed and fabricated the spent fuel racks for Hope Creek.
The quality assurance program utilized by par to assure that the Boral is securely encap-sulated into the stainless steel wall of the specified storage cell in accordance with storage rack design is par-OCP-64-9028 entitled
" Inspection Procedure for Square Stainless Steel Tubes, Cavity Weldments, and Module Assembly."
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The procedure establishes guide-lines for traceability of material i
and fabrication and inspection of welds.
This includes welds on the wrapper, the boral containing portion of the spent fuel rack, a
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6c Record of the material, welder, inspection, and needed repairs is made on inspection record sheets that are a part of the P AR procedure.
One modified storage rack has the capability of storing fuel assem-blies and 14 defective fuel bundles.
The spent fuel racks are designed to handle irradiated or unieradia-ted fuel assemblies.
The shielding for the stored spent fuel assem-blies is designed to protect plant i
personnel from exposure to direct i
radiation greater than that permitted for continuous occupa-tional exposure during normal operations.
The center-to-center spacing for the fuel assembly between rows is 6.308 inches.
The center-to-center spacing within rows is 6.308 inches.
Fuel assembly placement between rows is not possible.
Lead-in and lead-out guides at the top of the racks provide guidance of the fuel assembly during inser-tion or withdrawal.
The spent fuel storage racks are designed to withstand a pullup force of 4,000 lb..and a horizontal force of 1,000 lb.
There are no readily available forces in excess of 1,000 lb.
The racks are designed with lead-outs to prevent sticking.
However, in the event of M P8 4 153/01 ll-az
7 a stuck fuel assembly, the maximum lifting force of the refueling platform grapple (assuming limit switches fail) is 3,000 lb.
A complete description of the spent fuel storage racks is contained in Section 9.1.2 and Appendix 9B of the HCGS FSAR.
1.2.4.2 Fuel Handling System All required fuel handling equip-ment will be preoperationally tested for safe operation prior to its use for f uel handling activi-ties.
The fuel handling equipment and fuel bundles and assemblies are specifically designed for all fuel handling activities described in this application.
A complete description of the Fuel Handling System is contained in Subsection 9.1.4 of the Hope Creek FSAR.
1.2.4.3 Fual Handling Activities Upon arrival of a shipment of fuel the following will take place:
1.
When the new fuel delivery truck arrives on site, the Senior Nuclear Shif t Supervisor and Reactor Engineering Department representative will be notified.
Radiation Protection personnel will a
perform a radiation survey of the delivery truck.
The Senior Nuclear Shift Supervisor and Reactor Engineer will be notified of any unsatisfactory survey results.
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The shipment is then directed j
from the gate to the Railroad Access door located on the south face of the Reactor Building under escort of Radia-tion Protection personnel.
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3.
Maintenance personnel will locate the truck and direct the removal of tarps and chains.
4.
Radiation Protection personnel will survey the wooden crates.
a 5.
The shipment and shipping con-i tainers will be verified to
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comply with shipping papers presented by the carrier.
Reactor Engineering is respon-sible for evaluation and reso-j lution of discrepancies.
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6.
Upon proper acceptance of ship-
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ping papers and radiation sur-i veys, the truck may be unload-i ed.
If the shipping papers are t
incorrect, the truck may be unloaded, provided the contain-i ers are properly tagged and treated as nonconforming matarial.
1 The metal shipping containers will I
be removed from their outer wooden containers and hoisted to the 201' 1
elevation, of the Reactor Building I
using the Reactor Building Crane Auxiliary Hoist.
During removal of the metal shipping containers from the wooden shipping crates, Radia-tion Protection personnel will survey the metal containers.
The fuel may now be readied for inspec-tion, channeling, and storage or inspection and storage.
All personnel involved in the l
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inspection operation will be trained and will have reviewed the procedures for fuel receipt, handling, storage and criticality safety.
Inspection, channeling, and storage will proceed in accordance with written procedures as follows:
1.
Unpack fuel bundles from the metal shipping containers.
Remove the polyethylene sleeves from the fuel bundles prior to inspection.
i After the polyethylene sleeve is removed, Radiation Protec-tion personnel will perform a survey to ensure no external l
contamination is present.
The sleeves will then be perma-nently discarded.
2.
Move one bundle to the new fuel inspection stand and secure in place on the inspection stand.
Move second bundle from the shipping container and secure in place on the inspection 1
stand.
Two bundles may be secured on the inspection stand concurrently.
3.
The inspection will encompass the following categories:
a.
Visual examination b.
Removal of packing spacers 1
c.
Dimensional check d.
Pin enrichment and location check (also gadolinium fuel pins) e.
Clean all outside surfaces and verify cleanliness of all visible surfaces.
f M P84 153/01 14-az I.
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4.
The inspected bundles may now be channeled and transported to the spent fuel storage pool.
i 5.
In addition to the fuel handling platform operator, an l
l independent observer will verify the coordinates of the stored fuel in the spent fuel storage pool.
After the fuel has been stored dry in the spent fuel pool racks, the fuel will be j
covered until the pool is flooded for neutron source assembly and installation and fuel loading.
The fuel will be covered by placing 1
a tarp directly over the spent fuel storage racks containing the new fuel assemblies.
i This arrangement will not prevent water from draining from the assemblies in the event of j
flooding and subsequent draining of the. fuel storage area.
i Should a defective new fuel bundle be found, the bundle will be clearly marked and segregated from all non-defective fuel bundles in the spent fuel storage pool.
l Transfer of new fuel stored in the Railroad i
Access area at elevation 102' to the operat-ing deck at 201' elevation will be made as soon as possible.
Every ef fort will be made to minimize the time of storage.of new fuel in the temporary storage area.
1.2.5 Fire Protection System l.2.5.1 General Description t
The materials used in construction of the fuel storage area are 0
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i concrete and steel.
The fuel assemblies and fuel racks are also t
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constructed of non-combustible l
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materials.
Fire suppression i
equipment consists of manual water
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hose stations and portable fire i
extinguishers.
Details of the Fire l
Protection System are found in the HCGS FSAR Section 9.5.1, and Appendix 9A.
Reference drawings showing the relative location of all fire pro-tection apparatus (i.e., hose sta-tions extinguishers, etc.) in the Reactor Building are also shown in i
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The Fire Protection Program for the
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entire plant will not be complete j
and implemented at the time of the j
fuel delivery.
However, the Fire Protection Program for the areas where new fuel is stored and handled will be complete by the a
time of initial fuel receipt or appropriate mitigating measures which will provide an equivalent level of protection as defined in BTP CMEB - 9.5-1 will be provided.
As a minimum, the following Fire Protection features will be available at fuel recsipts i
a)
Fire hose stations in the Railroad Access area (Elevation 102') and Refueling Floor (Elevation 201') will be complete and preoperationally tested, including water supplies and pumps; i
M P84 153/01 16-az 4
c-lla b)
Portable fire extinguishers will be in place and functional; 1
c) 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> por day fire watch (security guard) and fire brigado coverage to the fire protection program; d)
Administrative procedures and training related to the fire protection program complete and implemented for the fuel storage area.
e)
Dedicated fire water storage of 75,00 gallons (6 hose stations for two hours).
The features should provide at least an equivalent level of fire protection for the areas related to fuel delivery as if the entire plant Fire Protection Program was implemented.
The Reactor Building Ventilation System (RBV3) will not be functional at fuel receipt.
However, temporary fans located at elevation 178' shall provide clean filtered air through temporary ducts to the refueling floor at elevation 201'.
The function of these fans is to aide in maintaining dust levels in the refueling floor area to acceptable levels.
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1.2.6 Access Controls A description of the controls for prevention of unauthorized access to the fuel storage areas is contained in the HCGS " Physical Security Plan for the Protection of Special Nuclear Material of Low Strategic Signifi-cance."
This plan is considered security confidential and as per the requirements of 10 CFR 73.21 must be withheld from unautho-rized disclosure.
This security plan is sub-mitted under separate cover.
1.3 PHYSICAL PROTECTION The quantity of U-235 (contained in uranium enriched to 20%, or more in the U-235 isotope), to be possessed under this license is less than the quantity specified 1
in 10CFR73.1(b) of 10CFR73.
Therefore, the physical j
protection requirements specified therein do not apply.
1.4 TRANSFER.OF SPECIAL NUCLEAR MATERIALS The General Electric Company, the fuel fabricator, is responsible for the shipment of the new fuel assem-blies from the fabrication plant at Wilmington, North Carolina to the Hope Creek plant site.
The fuel will be shipped in General Electric's model RA-2 or RA-3 containers, which are certified as fissile Class I containers by the most current revision of the NRC Certificate of Compliance USA /4986/AF and are autho-rized for the transport of fissile radioactive mate-rial in the form of General Electric reactor f uel.
As required by 10CFR70.51(c), Public Service Electric and Gas will establish, maintain, and follow written material control and accounting procedures sufficient t
to conduct the physical inventories required by 10 CFR 70.51(b), to maintain the accounting records required by 10CFR70.51(b), and to submit the material j
status reports and nuclear material transfer reports required by 10CFR70.53 and 10CFR70.54.
M P84 153/01 17-az
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In the event fuel must be returned to the GE facility, PSE&G will be responsible for proper packaging of fuel for return shipment.
All packaging of fuel by PSE&G for transport will be done in accordance with 10 CFR l
Part 71.
l.5 FINANCIAL PROTECTION AND INDEMNITY Because Public Service is an applicant other than a l
Federal agency or a nonprofit educational institution, Public Service comes under the requirements of Title 10 CFR, Part 140, Subpart B, Section 140.13, which requires a holder of a construction permit, who is d
also a holder of a license under 10 CFR Part 70,
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authorizing ownership, possession, and storage of 4
Special Nuclear Material, to have and maintain finan-cial protection in the amount of S1,000,000.
Proof of i
financial protection should meet the requirements of 10 CFR Part 140, Section 140.15.
t j
Public Service intends to obtain a policy of liability insurance in the amount of $1,000,000 which is an acceptable form of financial protection as stated in 10 CFR Part 140, Section 140.14, " Types of Financial l
Protection."
The policy will be ef f ective prior to receipt of fuel at the Hope Creek site.
Proof of
.I financial protection will be supplied to the NRC as a copy of the liability policy, together with a certifi-cate of authenticity provided by the insurer, as pro-vided by 10 CFR Part 140, Section 140.15(a)(1).
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2.0 HEALTH AND SAFETY 2.1 RADIATION CONTROL j
2.1.1 Oualifications l
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The technical qualifications for personnel
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j, with Radiation Protection responsibilities 1
are described in FSAR Section 13.1.3 and I
sections referenced therein.
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j 2.1.2 Responsibilities The responsibilities of key Radiation Protec-
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tion personnel are described in Section i
12.5.1.1 of the FSAR.
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!j 2.1. 3 Training and Experience The training and experience of Radiation Pro-tection personnel is described in FSAR Sections 13.1.3, 13.2.1, and 12.5.1.2.
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2.1.4 Contamination Monitorinq
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Radiation and contamination monitoring will
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be performed prior to the initial handling and storage of new fuel.
All new fuel that j
has not been unloaded or unpacked will be i
handled as contaminated material with all
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appropriate radiological controls in effect until contamin.a. tion checks are performed.
New fuel will be checked for radioactive con-tamination by Radiation Protection personnel j
as part of the new fuel inspection proce-1 dure.
Swipes or smears will be taken of the i
fuel in order to obtain a representative 1
sample of the surface contamination of the i
entire assembly and will be counted for alpha I
and beta / gamma activity to determine the amount of contamination present.
If the amount of contamination is found to exceed I
allowable limits, the source of contamination i
will be determined and appropriate decontami-j nation steps will be initiated as required.
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15 The Radiation Protection program outlined in FSAR Section 12.5 describes the procedures and equipment involved in radiological controls.
There should be no significant radiation hazards associated with the unieradiated fuel and the handling and storage of the fuel as outlined above should be suf ficient to main-tain radiation exposures ALARA.
2.1.5 Instrument Calibration Instrumentation for detecting and measuring radiation consists of counting room equip-ment, portable instrumentation, and air samplers.
Capabilities for detecting alpha, J
beta, gamma, and neutron radiation are pro-vided.
Sufficient inventory is provided to accommodate use, repairs, and calibration.
Details of the onsite instrument calibration capabilities, including sources, equipment, and methods are not available to date, con-l sidering the calibration f acility is in the early planning stage.
Onsite calibration capability details will be available by 4
July 1,1985. The projected numbers and types i
of portable survey instruments and equipment to be utilized are presented in HCGS FSAR Table 12.5-1.
As the planning progresses, additional details of the calibration program will be included in Section 12.5.2.2 of the HCGS FSAR.
All of the planning will incorporate standard health physics practices and the recommendations of the appropriate regula-tions and publications.
Sufficient chemical supplies, chemistry laboratory equipment, and analytical instru-ments are available to perform the required sample preparations and analyses in support of radiation protection functions.
M P84 153/01 20-az
16 2.1.6 Conformance to 10CFR20 The Radiation Protection program consists of policies, procedures, instructions, rules and practices to keep individual radiation expo-sure within the limits set forth in 10CFR Part 20 " Standards for Protection Against Radiation" and to maintain total radiation exposure of personnel as low as is reasonably achievable (ALARA).
The program assures that Radiation Protection training is provided, that personnel and in-plant area radiation monitoring is per-formed, that records of training, exposure of personnel and surveys are maintained and that proper instrumentation is available and properly calibrated.
The Radiation Protec-tion Program is discussed in Section 12.5 of the FSAR.
2.1.7 Disposal of Wastes Any radioactive waste generated in relation to material contained in the license applica-tion will be stored on site until authorized for disposal at a commercial waste disposal facility.
2.2 NUCLEAR CRITICALITY SAFETT 2.2.1 Key Personnel Qualifications The Technical Engineer and the Senior Reactor Supervisor are the key personnel having nuclear criticality safety and fuel handling responsibilities.
The minimum qualifications for these key personnel are in accordance with Regulatory Guide 1.8 " Personnel Select-ion and Training."
The Technical Engineer and Senior Reactor Supervisor correspond to Technical Manager and Reactor Engineering positions respectively, contained in ANSI /ANS-3.1 as endorsed in Regulatory Guide 1.8 Refer to FSAR Section 1.8.1.8 and Table 13.1-3 for additional information.
M P84 153/01 21-az
a 17 2.2.2 Key Personnel Responsibilites The responsibilities of the Technical Engi-neer are as follows:
Technical Engineer - The Technical Engineer is responsible for the areas of reactor engineering, technical reports and procedures, thermal performance, equipment reliability monitoring and testing, and document control.
Reporting to the Trchnical Engineer are the Senior Reactor Supervisor, Senior Engineer-Technical and the Senior Engineer-Technical Staff.
The Senior Reactor Supervisor assumes authority and 1
responsibility in his absence.
l The responsibilities of the Senior Reactor Supervisor are as follows:
Senior Reactor Supervisor - The Senior Reactor Supervisor is responsible for Reactor Engineering and Thermal Performance and equipment reliability monitoring.
Engineers are assigned to the senior reactor supervisor to develop and implement the details of the programs.
The reactor engineering group assists the principal startup engineer in the development and implementation of initial criticality, low power physics and power ascension test programs and provides technical direction to the operations for thermal and nuclear operation of the reactor and initial core loading and refueling operations.
The reactor engineering group also monitor, collect, trend, and analyze performance data for systems important to plant efficiency and reliability.
2.2.3 Storace of Loaded Shipping Containers Fuol bundles may be stored temporarily in shipping containers.
If they are stored in this way, the shipping containers will be 4
M P84 153/01 22-az
18 stored in an array which is no more act ive than the array used during shipping.
Con-tainers will be stacked no more than three containers high when fuel bundles are con-tained within.
Array safety is based upon analyses performed by General Electric and presented in the General Electric SNM License No. 1097, Docket 70-1113, Revision 3, dated May 14,198 4.
This license was approved by the NRC Division of Fuel Cycle and Material a
Safety dated June 29, 1984.
Section 1.8.4.3 of the General Electric SNM license states
" Arrays can be constructed without limit to the number of containers so stored, except i
that each array shall be stacked to a height l
of no more than four containers high with each container separated by nominal 2 inch I
wooden studs and with the width and length for each array and separation between arrays determined only by container handling requirements."
Shipping containers will be located in limited access areas on the 201' elevation operating deck.
The fuel bundles are shipped in a. steel l
container (179 1/2" x 17 7/8" x 11") encased in a wooden shipping crate (206 3/4" x 29 j
3/4" x 31").
One (1) steel container is con-l tained in each wooden shipping crate.
Two (2) fuel bundles are contained in each steel
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j container.
The container and crate are described in General Electric Company drawing
{
numbers 769E321, 769E23 2, and 769E229.
2.2.4 Criticality Control / Spent Fuel Pool The design of the spent fuel storage racks provides for a suberitical multiplication j
factor (keff) for both normal and abnormal storage conditions.
For normal and abnormal i
conditions, keff is equal to or less than 0.95.
Normal conditions exist when the fuel storage racks are located in the pool and are i
l M P84 153/01 23-az l
i t
j v
.....,._-_r,_
~
19 covered with a depth of water approximately 25 feet above the stored fuel for radiation shielding and with the maximum number of fuel assemblies or bundles in their design storage position.
An abnormal condition may result from accidental dropping of a fuel assembly or damage caused by the horizontal movement of fuel handling equipment without first disengaging the fuel from the hoisting equipment.
The spent fuel storage array is such that keff is less than 0.95 due to the presence of the neutron absorber material which is I
attached to the rack structure.
The design of the f uel, racks, and pools ensures that water will not be retained around an assembly when the pools are flooded and then drained.
The racks are designed to maintain a fuel spacing of 6.308 inches (center-to-center) within a rack module.
i Neutron poison is used in the spent fuel racks.
No credit is taken for burnable 4
l poisons which may be contained in any fuel bundles.
For additional information on Spent Fuel Pool refer to HCGS FSAR Section 9.1.2.
The safety evaluation of the Spent Fuel Pool is provided in Subsection 9.1.2.3 of the Hope Creek FSAR.
Criticality analysis is presented in subsection 9.1. 2.3. 3.
Each fuel movement is required by procedure to be confirmed by an independent observer before the movement is considered complete.
2.2.5 Criticality Safety Based on Other Than Maximum Enrichment of Fuel This section of Regulatory Guide 3.15 is not applicable.
Criticality safety is based on new fuel with a nominal flat U-235 enrichment of 3.4 w/o.
For additional information refer M P84 153/01 24-az
,ry
~
20 to Section 2.2.4 of this document and Subsection 9.1.2.3.3 of the HCGS FSAR.
2.2.6 Criticality Safety Based on the Reactivity Effects of Neutron Absorber Materials Refer to Section 2.2.4 of this document and Subsection 9.1.2.3.3 of the HCGS FSAR.
2.2.7 Criticality Safety Based on Moderation Control Refer to Section 2.2.4 of this document and Subsection 9.1.2.3.3 of the HCGS FSAR.
2.2.8 Validation of Calculational Method for Criticality Safety Description of the computer codes and metho-dology utilized in the verification of the HCGS criticality analysis is presented in l
FS AR Section 9.1. 2. 3.3.
2.2.9 Maximum Number of Fuel Assemblies Out of Authorized Locations The maximum number of fuel assemblies that will be allowed outside a normal, approved f
storage location or normal shipping container is three (3).
Fuel assemblies outside approved storage locations or shipping con-tainers must maintain an edge-to-edge spacing of 12 inches or more from all other fuel.
A fuel array of four or more assemblies outside approved fuel storage locations or shipping containers is prohibited.
No more than one metal shipping container containing fuel may be opened at any one time, and this container must be closed if j
all fuel is not immediately removed.
)
l l
M P84 153/01 25-az
21 Removal of wooden crates is done in the Fuel l
Handling area at elevation 102'-0".
The metal shipping container will be opened only in the fuel handling area (fuel container opening area) at elevation 201'-0".
2.2.10 Request for Exemption Public Service Electric & Gas Company (PSE&G) requests exemption from the monitoring and emergency procedures requirements of 10CFR70.24.
This exemption is requested because of the nature of the special nuclear material storage arrangements and procedural controls which PSE&G proposes to employ precludes any possibility of accidental criticality during receipt, unloading, inspection, storage, or packaging of the new fuel assemblies.
2.3 ACCIDENT ANALYSIS 2.3.1 Fuel Building & Reactor Building Detailed accident analyses of fuel handling equipment and storage areas are provided in HCGS FS AR Sections 9.1. 2 and 9.1. 4.
The accidents considered that could affect the safety of new fuel in the fuel handling and storage area are as follows:
Railroad Access Area Dropping of a single container containing two fuel assemblies in the receiving area lifting bay while being lifted by the fuel building polar crane.
The consequences of this accident would be limited to impact damage to the dropped con-tainer and any container impacted in the Railroad Access area awaiting movement to the fuel container upending area.
Fuel damage from this accident would be limited to the possible rupture of fuel rods in the dropped M P84 153/01 26-az
22 and impacted containers.
Since this accident affects only new fuel the consequences would be limited to the potential release of unirradiated uranium dioxide fuel.
No potential for a criticality condition exists in this accident since the maximum number of containers is enveloped by the 10CFR71 analysis for the shipping containers.
Fuel Container Upending Area Dropping of a single container containing two fuel assemblies to the floor at the 201'-0" elevation of the Reactor Building or falling over of an upended and open fuel container.
The consequences of these handling accidents would be limited to impact damage to the dropped container or fuel assemblies.
Fuel damage from these accidents would be limited to the possible rupture of fuel rods in the dropped containers.
Since this accident affects only new fuel the consequences would be limited to the potential release of unirradiated uranium dioxide fuel.
No potential for a criticality condition exists in this accident since only one container containing at most two fuel assemblies is involved.
other Accidents All other handling accidents involve only one fuel assembly and are discussed in FSAR Sections 9.1.2 and 9.1.4.
No overhead load greater than one fuel assembly will be allowed over any fuel storage array or rack which contains new fuel.
The seismic design of the Reactor Building and of cranes, racks, and pools precludes the credibility of more severe accidents.
In the unlikely event of a dropped new fuel assembly in the storage areas, the consequences would be minimal.
Due to the spacing of M P84 153/01 27-az
t 23 storage arrays, a criticality condition would not be possible under these accident conditions.
The consequences of these accidents would be limited to the possible rupture of new fuel rods and subsequent release of unirradiated uranium dioxide fuel.
2.3.2 Temporary Storage Area To preclude damage from f alling objects no construction loads will be allowed over the fuel in the Railroad Access area.
l M P84 153/01 28-az
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TABLE 1.1.1-1 j
i PUBLIC SERVICE ELECTRIC AND GAS COMPANY Directors James R.
Cowan Kenneth C.
Rogers T. J.
Demot Dunphy Verdell L. Roundtree Robert R.
Ferguson, Jr.
Nilliam E.
Scott Irwin Lerner Robert I. Smith William E. Marfuggi Harold W.
Sonn Marilyn M.
Pfaltz Robert V. Van Fossan James C. Pitney Josh S. Weston Officers Harold W.
Sonn Chairman of the Board, President and Chief Executive Officer William E. Scott Senior Executive Vice President Everett L. Morris Executive Vice President - Finance Frederick W.
Schneider Executive Vice President - Operations Frederick R.
DeSanti Senior Vice President - Customer Operations Richard M.
Eckert Senior Vice President - Nuclear and Engineering Robert W.
Lockwood.
Senior Vice President - Administration Stephen A. Mallard Senior Vice President - Planning and Research j
James B.
Bandel, Jr.
Senior Vice President Donald A. Anderson Vice President - Computer Systems and Services Lawrence R.
Codey Vice President and Corporate Rate Counse]
M P84 159/05 1-mw
j TABLE 1.1.1-1 ;
PUBLIC SERVICE ELECTRIC AND GAS COMPANY Directors Officers Robert M.
Crockett Vice President - Fuel Supply Robert H.
Franklin Vice President - Public Relations Carroll D. James Vice President - Administrative Planning Charles E. Maginn, Jr.
Vice President - Human Resources Wallace A. Maginn Vice President and Treasurer Winthrop E. Mange, Jr.
Vice President - Corporate Services Thomas J. Martin Vice President - Engineering and Construction Parker C.
Peterman Vice President and Comptroller Lousi L.
Rizzi i
Vice President - Customer and Marketing Services i
Robert J.
Selbach Vice President - Transmission and Distribution j
R.
Edwin Selover Vice President and General Counsel Robert S. Smith 4
Vice President and Secretary l
Rudolph D.
Stys Vice President - System Planning Corbin A. McNeil Vice President - Nuclear Richard A. Uderitz Vice President - Production M P84 159/05 2-mw
TABLE 1.1.1-1 PUBLIC SERVICE ELECTRIC AND GAS COMPANY l
Directors l
Officers Edward J. Biggins, Jr.
Assistant Secretary Marion F. Reynolds Assistant Secretary Rondald J. Hornak Assistant Treasurer Linda M. Prial Assistant Treasurer Donald J. Wallace Assistant Treasurer M P84 159/05 3-mw
)
TABLE 1.1.1-2 ATLANTIC CITY ELECTRIC COMPANY Directors Eleanor S.
Daniel Mack C. Jones Richard M.
Dicke Irving K.
Kessler John D.
Feehan Madeline H. McWhinney Jos. Michael Galvin, Jr.
John M. Miner Gerald A.
Hale Frank H. Wheaton, Jr.
1 Matthew Holden, Jr.
Richard M. Wilson Officers l
John D.
Feehan Chairman of the Board, President and Chief Executive Officer j
Ernest D. Huggard Executive Vice President Frank J.
Ficadenti i
Senior Vice President - Engineering and Construction Jerrold L. Jacobs Senior Vice President - Operations Michael A. Jarrett Senior Vice President - Corporate Services David V.
Boney Vice President - Customer and Community Services John F.
Born Vice President - Electric Operations Thomas E.
Freeman Vice President - Human Resources Meredith I.
Harlacher, Jr.
Vice President - Engineering Brian'A. Parent Vice President and Treasurer Joseph G. Salomone Vice President - Controls M P84 159/05 4-mw
TABLE 1.1.1-2 ATLANTIC CITY ELECTRIC COMPANY Directors Officers Henry C. Schwemm, Jr.
Vice President - Production Martin R. Meyer Secretary and Assistant Treasurer Lance E.
Cooper Controller Joseph T.
Kelly, Jr.
Assistant Vice President - Operations and Assistant Secretary e
9 M P84 159/05 5-mw
w TABLE 1.1.2-1 HOPE CREEK GENERATING STATION GENERAL FUEL DATA FUEL ASSEMBLY DATA Number of fuel rods 62 Number of nonfueled water rods 2
Rod array (square) 8x8 Rod pitch, inch 0.640 Number of fuel spacers 7
Spacer material 2r-4 with Inconel springs FUEL ROD DATA Fuel material UO2 Pellet o.d.,
inch 0.410 Cladding material 2r-2 with 2r liner Cladding tube o.d.,
inch 0.483 Cladding tube wall thickness, inch 0.032 Active fuel length, inch 150.0 WATER ROD DATA Outside Diameter, in.
.591 l
Inside Diameter, in.
.531 IM:vw MP84 123/17 2-vw O
e
e TABLE 1.1.3-1 HOPE CREEK GNERATING STATION FUEL ASSEMHLY TYPES Average Burnable Poison Bundle Average ~
Uranitzn per Bundle Maximum Pin Maximisa Pin Bondle Number of Pin Enrichment Bundle U235 Enrictynent Enrichment Ntsnber Bundles (N/O U235)
(Kg)
(Kg)
(W/O U235) g o u235)
Rods (W/O Gd203) 1 92 0.711 183.000 1.301 0.711
- 0.711 0
0
- 2 132 0.94 182.985 1.720 1.20 1.16' O
O 3-160 1.63
-182.656 2.977 2.00 1.90-2 5.0 2
2.0 4
308 2.48 182.403 4.524 3.80 3.55 4
5.0
~
5
-72 2.78 182.660 5.078 3.60 3.37 3
3.0 10 DES 764 Fuel Bundles 139, 546.620 Kg Uranantsn 72531.968 Kg U235 IM2w MF64 123/17 1-w
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