ML20128E645
| ML20128E645 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek, 07003011  |
| Issue date: | 05/23/1985 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| 25332, NUDOCS 8505290436 | |
| Download: ML20128E645 (45) | |
Text
f
~
7 t
O PS G Company PutAC Serwce Electnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation May 23, 1985 Director of Nuclear Reactor Regulation
.U.S.
Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesdh, MD 20814 Attention:
Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:
REQUEST-FOR ADDITIONAL INFORMATION-SPECIAL NUCLEAR MATERIAL LICENSE HOPE CREEK GENERATING STATION DOCK ET NO. 50-354 Pursuant to the NRC Request for Additional Information (RAI) as presented in a letter from Kishore Kodali (NRC) to Bruce A.
Preston (PSE&G) dated April 5, 1985, Public Service Electric and Gas Company (PSE&G) submits the attached response-(Attachment I).
PSE&G notes that in completing responses to questions one through five of the subject RAI it was required that several pages of the initial Special Nuclear Material (SNM) submittal, as provided the NRC in a letter from R.
L.
Mittl (PSE&G) dated November 13, 1984, be amended.
The amended pages conveying the requested information are provided within the body of the revised SNM license application transmitted for NRC review via this correspondence
( Attachment II).
Revisions to the affected SNM pages are denoted by a bar in the right hand margin.
In addition, several editorial changes are incorporated into this revision of the SNM license application.
These editorial modifications are also denoted by a bar in the right hand margin.
PDR ADOCK 05000354 gO/
8505290436 850523 A
PDR The Energy People I
054312 (4M) 7 83
~
~
Dir'. Nucl. Reac. Reg.
2 5/23/85 Should-you'have any questions in this regard, please contact us.
Very truly - yours,
/
Attachment I - Response to RAI
- Attachment II - SNM License Rev.1
.C-D..H. Wagner USNRC Licensing Project Manager A. R. Blough USNRC Senior Resident Inspector I
i 1( P85 90/11 -1/2 o
l L-
7-ATTACHMENT I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION M P85 90/09 1
Question 1.
Page 6, Section 1.2.4.
a.
Describe the quality assurance program to assure that the Boral composition meets design specifications.
b.
Provide the composition of Boral, minimum w/o B4C, and the minimum Boral density.
c.
Describe the quality. assurance program 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.
Response
Page 6, Section 1.2.4 of the Hope Creek SNM license has been revised to provide the requested information.
M P85'90/09 2
Question 2.
Page 9, Section-l.2.4.3 Describe the type of covers to be.used to pro-tect.the fuel assemblies.from dust, etc.
If the assemblies _are covered,-state whether the covers are open at the bottom so that water would drain freely.from the assemblies in the event of flooding and subsequent draining of i
the. fuel storage areas.
Response' Page 10, Section 1.2.4.3 of the Hope Creek SNM
-license has been revised to provide the requested ~information.
MP 85 90 09/3
7.
Question 3.. Page 16, Section 2.2
-Describe the plans for compliance with the requirements of 10 CFR 70.24, including instrumentation, location of detectors, and emergency procedures and drills.
Response
Page 21.of the Hope Creek SNM license is amended and reflects the addition of.Section 2.2.10 " Request for Exemption."
This section describes the plans for compliance with the requirements of 10CFR70.24.'
)
s a
l' l
MP 85 90 09/4-bp i
1 l-
f
.:T (l[
'4-
\\
. 0uestion'4.
Page 16, Section 2.2.1
)'.
Identify the Technical Engineer and' the Senior Reactor-Supervisor.with.the related positions in Regulatory Guide 1.8.-
Response
Page 16,_Section 2.2.1 of the Hope Creek SNM license has been revised to provide the requested information.
/
b t
- MP 85 90 09/5-bp e
s
--.e y,,
-.e
,-e.
r-sr w r,
, -.. - - ~,.
m-,,m--
Question 5.
Page 18, Section 2.2.3 Identify the " shipping container" (e.g.,
inner metal and outer wooden or inner metal only) with the related array size and shape (e.g.,
infinite aray, 3 container high array) and provide the basis for their safety.
Response
.As stated in the Hope Creek Special Nuclear Material (SNM) license application Section 2.2.3, "The fuel bundles are shipped in a. steel container - (179-1/2" x 17-7/8" x 11") encased in a wooden shipping crate (206-3/4" x 29-3/4" x 31").
One (1) steel container is contained in each wooden shipping crate.
Two (2) fuel bundles are contained in each steel container.
The container and crate are described in General Elecric Company drawing numbers 767E321, 769E232, and 769E229."
For additional information refer to SNM license Section 1.4 which identifies the shipping containers as General Electric model RA-2 or RA-3 containers, which are certified as fissile Class I con-tainers by the most current revision of the NRC Certificate of Compliance USA /4986/AF and are authorized for the transport of fissile radio-active material in the form of General Electric reactor fuel.
With respect to related array size and shape, the Hope Creek SNM license section 2.2.3 states that if fuel bundles are stored in their ship-ping container the storage array will be no more active than the array used during the actual shipping.
Containers will be stacked no more than three containers high when fuel bundles are contained within.
Shipping con-tainers will be located in limited access areas on the 201' elevation operating deck.
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, 1984.
This license was approved by the NRC Division of Fuel Cycle and Material 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 that each array shall be stacked to a height of no more than four containers
'M P85-90/09 6-bp
- Response : (Cont ' d ) -
-high with each container separated by nominal 2
' inch wooden studs and with the width-and : length for'each array and. separation between arrays determined only-by container handling requirements."
Page 18, Section 2.2.3 of the Hope Creek SNM license is amended to reflect the requested information, specifically.with respect'to array safety.
L h
s F
[
i L
[
MP 85 90 09/7-bp E
/
L'-
Ouestion 6.
FSAR, Page 9.1-12 Provide the basis for safety of a fuel assembly along the outside of the storage array (Section
-9.1.2.3.1, d).
Response
The basis-for safety of a fuel assembly along the'outside of the storage array is provided in Hope Creek FSAR Section 9.1.2.3.3.4.2.b.
. ;i
(
(.4 c.
MP 85.90 09/8-bp l
l k
Question-7.
FSAR, Page 9.1-22 Specify the controls that prevent insertion of assemblies-in the non-borated 5x6 special rack
( Fig 9.1-41 ).
The safety of fuel assemblies in this array has not been evaluated.
Response
The controls that prevent insertion of assem-blies in the non-borated 5x6 special fuel rack are specified in the Hope Creek Special Nuclear Material License (SNM) Section 1.2.4.3 and require that an independent observer verify the coordinates of the stored fuel in the spent fuel storage pool during and after fuel movement operations.
With respect to the safety of fuel assemblies stored in the non-borated 5x6 special rack, FSAR section 9.1.2.3.3.6.1 states that the.
storage of undamaged fuel within the special rack is less reactive than storage of damaged fuel.
This is due to the fact that in the ruptured fuel case, the defective fuel storage container displaces water.
For this reason the storage of undamaged fuel was not analyzed.
This conclusion is based upon analyses of the storage of ruptured fuel assemblies within defective fuel storage containers inserted into the special rack using the CHEETAH-B/PDQ-7 diffusion theory model.
The case was analyzed
-as an infinite array in order to simulate storage of 30 ruptured fuel assemblies in the special rack.
The resulting Kegg for this case was-0.6589.
Considering this Keff accounts for no radial or axial leakage, the reactivity for the storage of fuel in the special rack is well below the design limit Kegg of 0.95.
7 MP 8 5 9 0 09/9-bp -
g.
Question 8.
FSAR, Figure 9.1-3a Clarify whether the array shown in Figure 9.1-3a is used to store new fuel assemblies.
If they are, provide a complete description'and the basis for its safety.
Response
Figure 9.1-3a, b, and c were presented as part of the originally issued Hope Creek FSAR submittal dated March 1983.
All references to these figures have been deleted from the Hope Creek ~FSAR as of Amendment ~7 submitted in August 198 4.
The subject figures'have inadvertently not been deleted.
PSE&G will delete these figures from the Hope Creek FSAR via Amendment 11 submittal which is scheduled for issuance in July 1985.
MP 85 90 09/10-bp
ATTACHMENT II SPECIAL NUCLEAR MATERIAL LICENSE - REVISION 1 MP84 153/01 1
- HOPE CREEK GENERATING STATION Application for License For Storage Only of Unirradiated-Nuclear Fuel Revision 1
'Public Service-Electric and Gas Company, pursuant to Title 10, Code of Federal. Regulations, Part 70, hereby applies for a license to permit the receipt, possession, inspection, and storage of special nuclear materials in the
. form of unirradiated nuclear fuel.
l-
'MP8 4 153/01 2-mw
o -
e A
s 1
CHAPTER 'l
' GENERIAL INFORMATION
~
1.1 -
. REACTOR'AND FUEL 1.1.1.
Identification ~of Reactor, Parties to the
- Licensing Action, Geographic Location, Docket and: Construction Permit Numbers
-The application forLSpecial Nuclear Material (SNM) License is submitted by the Public Service Electric and Gas Company - (PSE&G) for the nuclearapower 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 -@ 100% steam flow, with gross electrical output of approximately. lll8 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-reference-provided.therein.
PSE&G is incorporated in the State of New Jersey-with its' principal office located at i.
80 Park Plaza, Newark, New-Jersey 07101..
f
-All directors and principal officers names are listed in Table'l.l.1-1, All are U
'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-E 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.
To the test knowledge of the applicant Atlantic-City Electric Company is not' owned b
or controlled by any alien, foreign corpora-tion or foreign government.
MP84_153/l'03-az i
I I
'l 2
The HCGS.isflocatedion 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-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 Fuel Assemblies Leach 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 threadedJends: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 issa 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 Inconel-X '
spring.. The fuel rodEspacers laterally sup-port the bundle rods, maintain rod spacing.
and geometry, as well as dampen any flow induced vibrations.
M_P84 153/01-4-az
v 2
n E
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-t'o 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-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 l
U-233, plutonium, depleted uranium, or
-thor i um.
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 Based upon the-data presented in Table..
1.1.3-1, a total of 764 assemblies containing
- approximately ' 258 2 kg of U-235 will be
~
received.
As such, Public Service Electric and Gas Company hereby requests a license for U-T t
4 H
.. M~P841153/01 5-az 4
.,..- ~ -, -.. -., -
2-~
m,.w.,
,-e.,_,,,,,._.-.
.., -. ~,,,....
mm... E,
_.,-em_..,.~._,.
g.-.,,,,,..
,. ~
4
'3000 kg of U-235 to allow for manufacturing tolerances and for the receipt of spare assemblies if required.
No licensing request is submitted for U-233,
-plutonium, depleted uranium or thorium.
1.2 STORAGE CONDITIONS l~.2.1 Storage 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 Storage Environment The fuel will be stored channeled or 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 operations other than fuel and component inspection, handling, and storage will be M P84 153/01 6-az
A 5
-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 fuel.
Loaded fuel shipping _ con-tainers or properly: designed overload test weights may be handled. in these areas pro-vided that they are at no time _ suspended over
'the fuel arrays in storage.
Any-non-fuel related activities which' must be conducted in the fuel handling area will be reviewed and approved by the Shif t Supervisor,. or his designee.
Any non-fuel activities in the Railroad Access area on elevation 102'-0" of the Reac-tor-Building will be restricted as_follows during fuel handling:
a.
No' painting, grinding, sandblasting, or similar activities are allowed.
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 affect' fire protection in the fuel handling area are allowed.
When fuel handling activities are not in progres's, 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.
j 1.2.4 Description of Storage Facilities 1.2.4.1 Spent Fuel Pool -- Reactor Building n
The spent fuel pool in the Reactor-
[
Building contains a storage space sufficient for 4084 fuel assemblies.
- M P84 153/01 7-az l '
k.
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 accordance with the OA requirements of 10 CFR 50, Appendix B.
Brooks & Perkins Corp. manufactures the Boral utilized by PSE&G in the Hope Creek Generating Station Spent Fuel Racks.
Assurance that-the Boral composition meets design specifications is achieved through use of B&P's procedure BP-100530AP.
M P84 153/01 8-az
1 y
6a Material traceability ~ is maintained for the aluminum skins, 'the boron
- carbide and the aluminum powder from the raw material stage through 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-inum.
The central layer-contains a uniform aggregate of fine boron 4
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.l.
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.
- M P84 153/01 9-az
6b Table 1 Chemical Composition - Aluminum (1100 Alloy) 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 %
r Total Boron 70.0 min.
g 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.
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-QCP-64-9028 entitled
" Inspection Procedure for Square Stainless Steel Tubes, Cavity Weldments, and Module Assembly."
The procedure establishes guide-lines for traceability of material and fabrication and inspection of welds.
This includes welds on the wrapper, the boral containing portion of the spent fuel rack.
M P84 153/01 10-az
r-6c Record of the material, welder, inspection, and=needed repairs is made on inspection record sheets that are a part of the PAR 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 unirradia-ted fuel assemblies.
The shielding for the stored spent fuel assem-blies is designed to protect plant personnel from exposure to direct 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-P84 153/01 ll-az
m s
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 fuel 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.
J 1.2.4.3 Fuel 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-l Protection personnel will perform a radiation survey of the delivery truck'.
The Senior
- i Nuclear Shift Supervisor and Reactor Engineer will be notified of any unsatisfactory.
survey results.
M P84 153/01 12-az
8-2 The shipment is then directed from the gate to the-Railroad Access door located on the south face of the Reactor Building under escort of Radia-tion Protection personnel.
3.
Maintenance personnel will locate the truck and direct the removal of tarps and chains.
4.
Radiation Protection personnel will survey the wooden crates.
5.
The shipment and shipping con-tainers will be verified to comply with shipping papers presented by the carrier.
Reactor Engineering is respon-sible for evaluation and reso-lution of discrepancies.
6.
Upon proper acceptance of ship-ping papers and radiation sur-veys, the truck may be unload-ed.
If the shipping papers are incorrect, the truck may be unloaded, provided the contain-ers are properly tagged and treated as nonconforming material.
The metal shipping containers will be removed from their outer wooden containers and hoisted to the 201' elevation, of the Reactor Building using the Reactor Building Crane Auxiliary Holst.
During removal of the metal shipping containers from the wooden shipping crates, Radia-tion Protection personnel'will survey the metal containers.
The f uel may now be readied for inspec-tion, channeling, and storage or inspection and storage.
All personnel involved in the M P84 153/01 13-az
9
~
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:
l ~.
Unpack fuel bundles from the metal shipping containers.
Remove the polyethylene sleeves from the fuel bundles prior to inspection.
~
After the polyethylene sleeve-is removed, Radiation Protec-tion personnel will perform a survey to ensure no external n
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 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 c.
Dimensional check d.
Pin enrichment 1and location check-(also gadolinium fuel pins) e.
Clean all outside surfaces and verify cleanliness of-all visible surfaces.
LM P84 153/01 14-az
10 4.
The inspected bundles may now be channeled and transported to the spent fuel storage pool.
5.
In addition to the fuel handling platform operator, an 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 covered until the pool is flooded for neutron source assembly and installation and fuel loading.
The fuel will be covered by placing a tarp directly over the spent fuel storage racks containing the new fuel assemblies.
This arrangement will not prevent water from draining from the assemblies in the event of flooding and subsequent draining of the fuel storage area.
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.
Transfer of new fuel stored in the Railroad 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 1.2.5.1 General Description The materials used in construction of the fuel storage area are M P84 153/01 15-az
-.-- ---a--
11 concrete and steel.
The fuel assemblies and fuel racks are also constructed of non-flammable mate-rials.
Fire suppression equipment consists of manual water type hose stations and portable fire extin-guishers.
All ventilation ducts penetrating fire barriers are pro-vided with rated fire dampers of the same rating as the barrier with 165* and/or 212*F fusible links.
Details of the-Fire Protection System are found in the HCGS FSAR Section 9.5.1.
Ventilation systems are discussed in detail in Section 9.4 of the Hope Creek Generating Station FSAR.
FSAR Section 9.5.1 discusses smoke removal and room isolation, Appendix 9 A, Table 9A-1 contains the fire hazards analysis summary.
Reference drawings showing the relative location of all fire pro-tection apparatus (i.e., hose sta-tions extinguishers, etc.) in the Reactor Building and Railroad Access ' area are shown in HCGS FSAR Section 9.5.1.
Construction of the Fire Protection System serving the Reactor Building elevation 201' and storage areas will be completed prior to the receipt of unirradiated fuel.
The Fire Protection System in these areas will also have successfully completed preoperational testing.
M P8 4 153/01 16-az
-t 12 1.2.6 Access' Controls x
-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-y mitted under separate cover.
'l.3 PHYSICAL PROTECTION The quantity of U-235 (contained in uranium enriched Lto 20%, or more in the U-235, isotope), to be possessed under-this; license is less than the quantity specified' in.10CFR73.1(b) of 10CFR73.
Therefore, the physical 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 tio the Hope Creek plant site.
The. fuel will:
be ~ shipped in General Electric's model RA-2 or RA-34 containers, which are~ certified as fissile Class I
. containers.by the most current revision of the NRC Certificate of 'Compli~ nce USA /4986/AF and are autho-a rized for the transport of. fissile radioactive mate-s rial in the form of-General Electric reactor fuel.
As required by 10CFR70.51(c),' Public Service Electric
.and Gas'.will establish,-maintain, and follow written-material control and accounting procedures sufficient to conduct theLphysical inventories required by
- 10 CFR.70.51(b), to maintain the accounting records
-required by 10CFR70.51(b), and to submit the material status reports'and nuclear material-transfer reports required by 10CFR70.53 and'10CFR70.'54.
M P84 153/01 17-az
I s
t 13 In -the event fuel must be returned ' to the GE facility, PSE&G will be responsible for proper packaging of fuel ifor--return shipment.
All packaging of fuel by PSE&G for - transport will be done in accordance =with 10 CFR E
Part 71.
1.5 FINANCIAL PROTECTION AND INDEMNITY:
Because'Public. Service is an applicant other than a 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 also a holder of a -license under.10 CFR Part 70, authorizing ownership, possession, and storage of Special Nuclear-Material, to have and maintain finan-cial protection in the amount of $1,000,000.
Proof of financial protection should meet-the requirements of 10 CFR Part-140, Section 140.15.
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
. Protection."
The policy will be.ef f ective prior - to receipt of fuel at the Hope Creek site.
Proof of 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).
3 M P84-153/01 18-az
I u-14
.2.0 HEALTH AND SAFETY 2.1;.
RADIATION: CONTROL ~
'2.1.1 Q ualifications -
The technical' qualifications for personnel with Radiation Protection responsibilities are described in-FSAR Section 13.1.3 and sections referenced therein.
2.1.2 Responsibilities-The responsibilities of key Radiation Protec-tion personnel are described in Section 12.5.1.1 of the FSAR.
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.
2.1.4 Contamination Monitoring Radiation and contamination monitoring will be performed' prior to the initial handling and storage of new fuel.
All new fuel that has not' been unloaded or unpacked will be handled.as contaminated material with'all appropriate radiological' controls in effect until contamination checks are performed.~
New fuel will be checked for radioactive con-tamination by Radiation Protection personnel as part of the new fuel inspection proce-dure. ' Swipes or smears will be taken of the f uel in order to obtain a representative sample of the surface contamination of the entire assembly and will be counted for alpha and' beta / gamma activity to determine the amount of contamination present.
If the amount of contamination is found to exceed allowable limits, the source of contamination will be determined and appropriate decontami-nation steps will be initiated as required.
M P84 153/01'19-az a
~
l 15
.The Radiation Protection program outlined in FSAR Section 12.5 describes the procedures -
and equipment involved in radiological controls.
j
-There should be no 'significant radiation l
hazards associated with the unirradiated fuel.
and the handling and storage of the fuel as j
outlined above should be suf ficient to main-I tain radiation exposures ALARA.
l l
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, l
beta, gamma, and neutron radiation are pro-vided.
Sufficient inventory.is provided-to I
accommodate use, repairs, and calibration.
Details of the onsite instrument calibration capabilities,-including sources, equipment,.
and methods are not available to date, con-I sidering the calibration facility is.in the early planning stage.
Onsite' calibration capability details will be available by j
July 1,1985. The projected numbers and types of portable survey instruments and equipment-i
.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 i
l
F e
16 e
~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. l~. 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 SAFETY 2.2.1 Key Personnel Oualifications 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 'l.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 L
1.8.
Refer to FSAR Section 1.8.1.8 and Table l
13.1-3 for additional information.
M P84 153/01 21-az i.
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 Technical Engineer are the Senior Reactor Supervisor,.
Senior Engineer-Technical and the Senior Engineer-Technical Staff.
The Senior Reactor Supervisor assumes authority and responsibility in his absence.
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 Storage of Loaded Shipping Containers Puel bundles may be stored temporarily in shipping containers.
If they are stored in this way, the shipping containers will be M P84 153/01 22-az
c 18 stored in an array which is no more active
-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.109 7, Docket 70-1113, Revision 3, dated May 14,1984
-This license was approved by the NRC Division'of Fuel Cycle and Material 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 that each array shall be stacked to a height of no more than four containers high with each container separated by nominal 2 inch 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 container (179 1/2" x 17 7/8" x 11") encased in a wooden shipping crate (206 3/4" x 29 3/4" x 31").
One (1)' steel container is con-tained in each wooden shipping crate.
Two (2) fuel bundles are contained in each steel container.
The container and crate are described in General Electric Company drawing numbers 769E321, 769 E23 2, - and 769 E229, 2.2.4 Criticality Control / Spent Fuel Pool The design of the spent fuel storage racks provides for a subcritical multiplication factor (keff) for both normal and abnormal storage conditions.
For normal and abnormal 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 M P84 153/01 23-az
e 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 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.
Neutron poison is used in the spent fuel racks.
No credit is taken for burnable 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 t --
f t
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 FSAR 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 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 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
-all fuel is not immediately removed.
1 M P84 153/01 25-az l
r r
b I-
n-21 Removal of wooden crates is done in the Fuel Handling area at elevation 201'-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 10CPR70.24.
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 1
Detailed accident analyses of. fuel handling equipment and storage areas are provided in HCGS FSAR Sections 9.1.2 and 9.1.4.
.The 4
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 lif ted by the fuel building polar crane.
The' con =aguances of-this accident would be limited to hapact 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.
t.
M.P84 153/0l ! 26-az ef 4
=:
m 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
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 falling objects no
- construction loads will be allowed over the fuel in the Railroad Access area.
s M P84 153/01 28-az
TABLE 1.1.1-1 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Directors
' James R.
Cowan Kenneth C.
Rogers T..J.
Demot Dunphy verdell L.
Round tree Robert R.
Ferguson, Jr.
William 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 James.B. Randel, Jr.
Senior Vice President Donald A. Anderson Vice President - Computer Systems and Services Lawrence'R. Codey
-Vice President and Corporate' Rate Counsel M-P84 159/05 1-mw
F-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 Vice President - Customer and Marketing Services Robert J.
Selbach Vice President - Transmission and Distribution R.
Edwin Selover Vice President and General Counsel Robert S. Smith Vice President and Secretary 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 Directors 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 I
TABLE-l.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.
fJohn - M. Miner Gerald A. Hale Frank H. Wheaton, Jr.
Matthew-Holden, Jr.
Richard M. Wilson Officers
-John D.
Feehan Chairman of the Board, President and Chief Executive Officer.
~ Huggard-Ernest D.
-Executive Vice President
. Frank J. Ficadenti
-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 1( P84 159/05 4-mw-
7.
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.
s h
1M'P84 159/05 5-mw
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 Zr-4 with Inconel springs FUEL ROD DATA Fuel material UO2 Pellet o.d.,
inch 0.410 Cladding material.
Zr-2 with Zr liner Cladding tube o.d.,
inch 0.483 Cladding tube wall thickness, inch- '
O.032 Active fuel length, inch 150.0 WATER ROD DATA Outside Diameter, in.
.591 Inside Diameter, in.
.531
-IM:vw-
.MP84 123/17 2-vw-
3 TABLE 1.1.3-1 HOPE CREEK GENERATING STATION.
FUEL ASSEMBLY TYPES Average Burnable Poison Bundle Average Uranitzn per Bundle Maximtzn Pin Maximum Pin Bundle Number of Pin Enrichment Bundle
1235 Enrichment Enrichment Ntznber Type Bundles (W/O U235)
(Kg)
' Kg)
(W/O U235)
(W/O U235)
Rods (W/O Gd203) s 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 0
0 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 5.078-3.00 3.37 3
3.0
'IUTALS 764 Fuel Bundles 139, 546.620 KQ Uraninum 2581.968 Kg U235 IMavw MP84 123/17 1-vw 1
4 t '
4'
-