ML19350C400
| ML19350C400 | |
| Person / Time | |
|---|---|
| Site: | 07002937 |
| Issue date: | 12/23/1980 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 18137, PLA-594, NUDOCS 8104010415 | |
| Download: ML19350C400 (28) | |
Text
M 7I'9 Two N O RT H NINTH STR E E T, AL L E N TOWN. PA. 18101 PH O N E s (2 iS) 8 21 5151 y
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Nuclear Material Safety and Safeguards c.
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U.S. Nuclear Regulatory Ccamission
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- Docket #50-387 SUSQ11El' ANNA-STEAM ELECTRIC STATION SPECIAL NUCLEAR MATERIALS LICENSE APPLICATh'i 8hg
- 70-29-37 Djg cn*
MA NEN FUEL STORAGE A'iD RECEIFT ER 100450 FILE 841
'PLA-594
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This application is filed pursuant to Title 10 Code of Federal Regulations,
.Part 70 for authori:ation to receive, possess, store, inspect, and package for transport nuclear fuel bundle / assemblies.* This application superceeds that which was filed on October 16, 1980 and has been revised to address concerns expressed by your Norm Ketzlach.
It is requested that the Special Nuclear Materials License remain in effect until receipt of the Operating License.
This application is excempt from licensing fees per 10CFR 170.11(3).
l Any_ questions concerning this application should be forwarded to Mr. Thomas E. Gangloff, (215)-821-5543 of the Nuclear Licensing Grcup.
The following information is submitted in support of the application.
1,0 APPLICANT Pennsylvania Power 6 Light Company P.O. Box 1S70 Allentown, PA.
18105 2.0 ADDRESS OF STORAGE SITE Unit 1 of the Susquehanna Steam Electric Station -(SSES) is located in Salem Township, Luzerne County, in-east central Pennsylvania, about
-five miles northeast'of.Berwick, Pennsylvania.
a
? ?lT*-A 3}[G
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' bundle - that which is received from the fuel manufacturer assembly - fuel bundle with channel affixed.
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3.0 CORPORATE INFORMATf0N The information set fcrth in the application for Construction Permit and Operating License, Docket No. 50-387 dated July 20, 1978 for the Susquehanna Steam Electric Station Unit 1, is hereby incorporated by reference.
Information concerning control and ownership of the applicant is also set forth in the application for Construction Permit
. and Operating License.
4.0 RADIOACTIVE MATERIAL l
Initial Core Fuel Assemblies c
Maximum square dimension of 5.47 in.
fuel' bundle / assembly Active fuel length 150 in.
Overall fuel bundle / assembly length 176.16 in.
Maximum square dimension of
- fuel channel-5.45 in.
Overall channel length 166.91 in.
Channel wall thickness
.080 in.
Rod. array type 8 x 8 = 64 rods 62 fuel rods 2 water rods 1
Fuel rod. Pitch' O.640 2n.
Fuel rod clad thickness 0.032 in.
Rod clad caterial Zircaloy Fuel rod outside diameter
.483 in.
Fuel pellet ~ diameter-
.483 in.
Fuel pellet material UO2 3
Nominal fuel stack density 10.32 gm/cm E
Total number of fuel 764 (3 types)_
assemblies Maximum U-235 enrichment (weight %)
3.00 Average.U-235 enrichment (reight-%)
1.88
~
Average U-235 enrichment-(weight %)
.711-1.76 2.19
- per assembly-type -
. Number of assemblies-92 240 432
. per. type--
2 f
t s - - - -. -. -- - - - - -.
---.---~? ^.
_,_Y
_ = _ _
e 00 weight per assembly (kg) 207.58 207.31 207.00 2
Gd 0 weight per assembly (kg) 0.0 0.20 0.44 Average UO Weight per fuel rod (kg) 3.34----------
i 2
U weight per assecbly (kg) 182.96 182.73 182.45 (1b) 403.36 402.85 402.23 The composition and location of each type fuel rod in the assembly
'having the highest enrichment can be fonnd in NEDC-21307, February 1977, entitled "BWR5 Standard Plant Fuel Design Summary" (pgs 15,17 &
18; figs 6 & 8). These figures are included as proprietary figures 4.3-3 and 4.3-5 of the_Susquehanna SES FSAR. Other fuel bundle types are described in NEDE 20944P.
The maximum amount of special nuc12ar material, as kilograms of U-235, which may be possessed at any one time is 3,043 kilograms. This 4
represents the initial core and fifty additional assemblies per enrichment type.
5.0 TRANSPORTATION J
All initial core fuel bundles are to be delivered to the site in accordance with shipping procedures and arrangements of the General Electric Company. The fuel hundles will be shipped by the fuel canufacturer as authorized under NRC Certificate of compliance number
- USA /4986/B(F) Revision 8.
Fuel will normally be shipped to the site by truck, with a maximum of 16 containers per truck and two bundles per container.
4 6.0 STORAGE. CONDITIONS
'The' fuei bundle / assemblies will be stored on the refueling floor of the Reactor Building (see SSES Final Safety Analysis Report, Section 9.1) in either the new fuel storage vault dry, in the spect fuel pool, i orcin their metal shipping containers on the refueling floor. When
<the fuel bundle / assemblies are storel in the spent fuel pool, the pool 3,
may: be dry, partially flooded, and/or completely flooded during -
residence in the storage rack.. The-refueling floor is surrounded by_
p'-
structural steel and metal siding and, therefore, it is considered unlikely that a fire will occur. :However,'the use of fla=mable materials ~ on the? refueling floor will belcarefully controlled as ant.
-added margin of safety.
In addition, the movement of the _new fuel vault cover-will be administrative 1y controlled by approved plant procedures..The refueling floor is shown on SSES Final Safety-
.AnalysisL eport (FSAR).FiguresLl.2-23,fl.2-24, 1.2-25, and;9.1-16B.
R
-Preoperational t'esting, to the extent ^ necessary-to prove that-the fuel handling equipment is functioning properly and is safe to handle the
'3 w-----
us.-
fuel over the fuel receipt, inspection and storage areas, will be performed prior to using this equipment to handle fuel. Likewise, the necessary support systems will be tested to the extent necessary to prove that they are capable of performing their specific functions for fuel receipt and inspection activities prior to the arrival of the first fuel shipment.
Functional checks of the fuel handling equipment will also be done in accordance with approved plant operating and surveillance procedures.
Some construction and preoperational testing activities may be occurring on the refueling floor (e.g. reactor well work, communication cable pulling, conduit installation) while fuel is Seing received, inspected, and/or stored. However, due to the security provisions that will be in effect, which include the separation of the two types of activities, there will be no impact on the safe receipt, inspection or storage of the fuel.
In addition, the responsible supervisor on shift during fuel receipt and inspection activities will have the authority to terminate either or both of fuel receipt and inspection activities or the construction and preoperational testing activities.
6.1 Inspection and Handling As soon as practicable after the fuel carrier vehicle arrives at the plant gate, it will be taken to a predetermined off load area l
where the shipping containers will be unloaded from the carrier vehicle. The wooden containers will then be taken to the i
unloading bay on the ground floor of the Reactor Building (EL672). The metal inner shipping containers will be removed from the wooden outer shipping containers. The metal containers, holding the fuel bundles, will be lifted to the refueling floor of the Reactor Building, where they can be stored for an interim period of time. The metal containers will be placed in a near vertical position, and the bundles are then removed from the containers. The fuel bundles may be placed in the new fuel inspection stand to be checked for integrity and numbering, or placed in an appropriate Storage location for future inspection.
Upon removal of the fuel bundle from the shipping containers, the plastic bag will be removed prior to inserting the fuel bundle into the storage rack.
There is no way water can be entrapped within the open lattice structure of the fuel bundle. No credible mechanism exists with which to block the 3-inch diameter nozzle on the bottom tie plate of.the fuel assembly.
All fuel' receipt and inspection activities will be controlled by the applicable procedures which are. reviewed and approved by the
~
Susquehanna SES Plant Operations Review Committee (See Section 13.4 of the Susquehanna SES FSAR), and approved by the Superintendent of the plant. The. ultimate responsibility for 4
administrative controls over fuel receipt and inspection activities lies with the Superintendent of Plant.
In the event of damaged or unacceptable fuel which must be returned to the manufacturer, packaging will be in accordance with DOT and 10CFR71 requirements and shipments will be made via a licensed carrier.
6.2 Safety Evaluation The new fuel bundle / assemblies can be stored in the new fuel storage racks in the new fuel storage vault, in high density spent fuel storage racks in the spent fuel pool and/or in their metal shipping containers. No more than three (3) fuel bundle / assemblies will be outside the shipping containers or storage racks at one time. The metal shipping containers are described, and their storage without a criticality warning system authorized in Amendment #5 to GE's Special Nuclear Haterials License #SNM-1097 dated June 6, 1978. When fuel bundles are stored in their metal shipping containers, the containers will not be stacked more than four (4) high. The cross-sectional size of the arrays will only be limited by space available on the refueling floor.
Metal shipping containers containing fuel will not be stored closer thar. f.*ve (5) feet to any fuel handling equipment. No interacti<,n will occur between fuel stored in the shipping containera and that stored in the new fuel storage vault as the vault is ewoeded in a concrete floor with a minimum of one (1) ft. of concrete between storage arrays.
Susquehanna SES Plant Staff Reactor Engineering personnel have the overall responsibility for nuclear criticality safety. These people are degreed physicists or nuclear engineers and have had training in BWR systems and fuel.
In addition they will be
-certified as fuel inspectors for Susquehanna SES and will receive all'the classroom training given to fuel handling or inspection personnel.
The criticality control and fuel rack-design criteria for the new fuel storage vault and the spent fuel storage pool is outlined as follows.
6.2.1
. Criticality Control - New Fuel Storage Vault Rack
.The new fuel storage vault is designed to maintain Keff
<.95 (see FSAR Subsection 9.1~.1.1.1.2) in dry c'onditions and when flooded with water between 0.7 and 1.0 g/cc. The new fuel vault is covered by leak proof, metal, removable covers which are controlled by 5-
administrative procedures which prevent optimum moderation conditions no more than two (2) vault covers will be off at a time.
The new fuel vault covers (FSAR Figure 9.1-la- ) overlap the storage vault curb and have a protective lip that prevents direct impingment upon the sealing point. The modified I-beams that span the vault also are overlapped 'ay the covers. The beams act as drains to direct run-r.ff from the covers to the refueling floor. Sealict is provided by the covers resting upon the channel and beam upper surfaces. The mass of the covers, and the -lose tolerances of construction, make gaps negligible.
The calculations of keff are based upon the geometrical arrangments of the fuel array and subcriticality does not depend upon the presence of neutron absorbing materials. To meet the requirements of General Design Criterion 62, geometrically-safe configurations of fuel stored in the new fuel array are employed to assure that keff will not exceed 0.95 if fuel is stored in the dry condition or if the abnormal condition of flooding (water with a density of 1 g/ce) occurs.
In the dry condition, Keff is maintained
<0 95 due to under-moderation. In the flooding and. flooded condition, the I
geometry of the fuel storage array assures the Keff will remain < 0.95 due to over-moderation.
No limitation is placed on the size of the new fuel storage vault rack array from a criticality standpoint since all calculations are performed on the basis of an infinite array of assemblies. The Nea Fuel Storage Vault will, therefore, accommodate fuel from a multi-unit facility with no safety implications. All handling conditions remain the same and there is no compromise of safety considerations.
6.2.2 New Fuel Storage Vault Rack Design a)
.The New Fuel Storage Vault contains 23 sets of castings which may contain up to 10 fuel bundle / assemblies; therefore a maximum of.230 fuel
. bundle / assemblies may be stored in the fuel vault.
b)
There are three tiers of castings which are positioned to fixed box beams. This holds the fuel bundle / assemblies in a' vertical position _and supports the fuel at the lower and upper t_ie plate 6
4 9
w r
with additional lateral support near the center of gravity of the fuel assembly, c)
The lower casting supports the weight of the fuel bundle /;--e nbly and restricts the lateral movement; the center and top casting restricts lateral movement only of the fuel burdle/ assembly.
d)
The New Fuel Storage Vault Racks are nade from aluminum. Materials used for construction are specified in accordance with ASTM-1978 specifications. The material choice is based on a consideration of the susceptibility of various metal combinations to electrochemical reaction.
When considering the susceptibility of metals to galvanic corrosion, aluminum and stainless steel are relatively close together insofar as their coupled potential is concerned. The use of stainless steel fasteners in aluminum to avoid detrimental galvanic corrosion is a recommended practice and has been used successfully for many years by the aluminum industry.
e)_
The minimum center-to-center spacing for the fuel bundle / assembly between rows is 11.875 inches.
The minimum center-to-center spacing within the rows is 6.535 inches. Fuel bundle / assembly i
placement between rows is not possible.
f)
Lead-in and lead-out of the casting provides guidance of the fuel bundle / assembly during insertion or withdrawal.
g)
The rack.is designed to withstand the impact force of 4000 ftlbs while maintaining the safe design basis..This impact force could be generated by the. vertical free fall of a fuel assembly from the height of 5.3. feet.
h)
The rack is designed to withstand the pull-up force of 4000 lbs. and a horizontal force of 1000 lbs. The racks are designed with lead outs to prevent sticking. However, in the event of a-stuck fuel bundle / assembly, the lifting bail will yield at a pull up force less than 1500 lbs.
1 i)
The rack'is designed to withstand horizontal combined loads _up to 222,000 lbs.,1well in excess of expected loads.-
7
. - - - =
j)
The maximum stress in the fully loaded rack in a faulted condition is 25.9 Kip. This is significantly lower than the allowable stress.
k)
The rack is designed to handle nonirradiated, low emission radioactive fuel assemblies.
6.2.3 Criticality Control - Storage of New Fuel in High pensity spent ruel Racks The design of the spent fuel pool storage racks assure that Keff will remain <0.95 for both normal and abnormal storage conditions. Normal conditions exist when the fuel is dry, flooding and flooded. An abnormal condition may result from accidental dropping of equipment or damage caused by the horizontal movement of fuel handling equipment without first disengaging the fuel from the hoisting equipment.
To ensure that the General Design Criteria 62 is met, the following normal and abnormal spent fuel storage conditions were analyzed.
a)-
normal positioning in the spent fuel storage
- array, b) fteel stored in control rod racks, c)
-Pool water temperature increases to 212 F, d) normal storage array of ruptured fuel, e) abnormal positioning in the. spent fuel storage
- array, f) moving fuel bundle.between work rack and storage
- area,
.g)
' dropped fuel bundle adjacent rack The criticality safety analysis was performed by Nuclear Associates International (NAI 78-75 dated 5 79). This analysis shows that under all postulated normal and. shnormal condit!ons, Keff will remain < 0.95.
A discussion of the analysis methodology and results
'follows:
6.2.3.1' The Principal-Analytical Model
-8
+
-~
The criticality analysis of the Susquehanna spent fuel storage rack design used the CEEETAH-B/CORC-BLADE /PDQ-7 calculational model as the basic engineering tool.
CEEETAH-B is the BWR lattice version of NAI's CHEETAH code which is a modified version of the original LEOPARD code and which uses a modified ENDF/B-II cross section library.
CORC-BLADE generates the equivalent diffusion theory cross-sections for the control blade.
The PDQ-7 program is the well known few group spatial diffusion theory code. The transport to d.f fusion theory correction is generated using the Monte Carlo model described in Section 6.2.3.2.
-The reference case storage rack cell is shown in Attachment 2.
A zero current boundary condition is applied to the 'our sides of this cell to produce an infinite array effect. The two dimensional PDQ-7 reference case calculations are made for four neutron energy groups, two mesh intervals per fuel pin, a flat U-235 enrichment description and a zero axial buckling to simulate infinite fuel length.
6.2.3.2 The Verification Model The verification calculation employs the KENO-IV/AMPX model. The basic neutron cross section data comes from the caster library.cf AMPX - a 123 group GAM-THERMOS neutron
. library prepared from ENDF/B version II data..
The NITAWL module of the AhPX program is used l-
'to. perform a Nordheim integral treatment of.
l the U-238 resonances accounting for the self-i shielding effect. The working library l
produced by the NITAWL/AMPX module retains L
.the 123 group energy structure and is used directly by KENO-IV.
In-the KENO-IV calculation, each fuel and-water rod cell is represented discretely, as l
a box type. The array "ption of KENO-IV is applied tc arrange the box types into a matrix representing the fuel assembly. Then-a water reflection region is added to the L
f outside of this matrix followed by th'e.
l; aluminum canister,. void aed Boral slab 9
t m_
regions to complete the reference case storage rack cell. To simulate the arrangement of a large number of storage rack units, and for a non-leakage condition in the axial direction, a specular reflective condition is applied to all six sides of the reference case storage rack cell (See ).
The KENO-IV/AMPX model has been benchmarked against earlier ORNL critical experiments measured by Dr. E. Johnson, a La Crosse Bh*R cold critical, and the more recent experimen* conducted by S.R. Bierman et. al.
at the Battelle Pacific Northwest Laboratory (PNL-2438). The latter results show the average calculated Keff to be 1.001+0.006.
6.2.3.3 Basic Assumptions The following assumptions were made in the analysis and ensure that the study follows a conservative approach.
1.
A uniform 3.25 w/o distribution in an 8-x 8 bundle (which is more reactive than a distributed enrichment with the same overall average weight percent).
2.
Fresh fuel, no burnable poison.
3.
Minor structural members replaced by water.
4.
Fresh water at 68*F.
5.
Fuel assemblies are channeled and centered within the storage cavity.
6.2.3.4 Effect on Reactivity of Uncertainties in Material Properties and Rack Geometry and Fuel Placement (a) Temperature Effect
-Using the nominal storage rack cell geometry, the temperature of the fuel and pool water was varied. Intermediate steps at 32*F, 95*F, 120 F, 150*F, 180*F, 212*F were, studied. The results of the analysis shows 10
that reactivity decreases continuously as temperature increases from 32 F.
(b) Void Effect The effect of boiling (assuming equat voids inside and outside of the rack) is studied by varying the voids from 0% to 20% at a temperature of 212 F with the nominal geometry. The results indicate a continuous decrease in reactivity for this range of voids.
(c) Boral Width Reduction The nominal width of the Boral slab is 5.250".
The change in reactivity due to the
-0.03 inch minimum tolerance on this width is AK=0.001.
(d) Channel Effect The rack must accommodate both channeled and unchanneled fuel. Studies reveal that the channeled fuel in the rack is more reactive than the unchanneled fuel. Taking the conservative approach, the study here involves channeled fuel except in the accident condition where unchanneled fuel is dropped. The decrease in ok from channeled to unchanneled fuel is 0.002 in the base case rack (See Attachment 2).
(e) Boral Acceptance Criterion The criticality calculations described in this report are based on the assumption that the core of the Boral slabs contains a homogeneous mixture of fine B C and aluminum 4
powder. An adjustment must be made to take into account directly the self-shielding effect due to the random distribution of B C 4
grains of~ finite sizes in the mixture.
Based on a homogeneous mixture, the neutron attenuation factor for a Boral slab of 80 mil thickness with 0.0232 g/cm2 B-10 loading was calculated to be 0.969. The vendor Brooks and Perkins has proposed the value 0.963 as the acceptance criterion for their measured 11 l
1
neutron attenuation factor for the above Boral slabs.
Based on NAI's analysis, the difference between an attenuation factor of 0.969 and 0.963 amounts to about 0.3% et for the Susquehanna Spent Fuel Pool. The reactivity reserve of the new pool is more than 1% Ak below the 0.950 limit permitted by NRC; hence Boral slabs meeting the 0.963 criteria will have no adverse effect on the safety of the pool.
(f) Change in Pitch (pitch sensitivity))
The recctivity effect of mechanical tolerance changes caused by structural, fabrication, installation, and seismic conditions can be conservatively determined by the pitch sensitivity study. The calculations were carried out on a 0.125 inch increment. The pitch reactivity coefficient was determined to be about'+0.7% Ak/0.125 inch pitch reduction in the range from 6.875 inch pitch to 6.375 inch pitch.
(g) Off-Ccnter Positioning The reactivity effect of off-center positioning due to improper loading or some natural phenomenon was investigated.
The
-calculations show that the reactivity decreases so no adverse effect occurs.
6.2.3.5 Results of the Criticality Analysis (a) Diffusion Theory Results -
The results of the criticality analysis of the proposed storage rack for Susquehanna Spent Fuel'are st=marized below:
LPDQ Results:-
.Keff Nominal Case 0.393 Dimensional'rtJ Positional Tole:2nce, AK 0.014-Boron Width Effect, AK-0.001
. Temperature Effect.,6K 0.004 Variation on Boral-Accept.
0.003 12 s-
Void Effect, AK Negative Adjusted Keff 0.915 (b) Monte Carlo Results and the Calculational Bias The KEN'J-IV verification calculation for the nominal case yields a Keff value of 0.912+0.004 with 95% confidence interval ranging from 0.904 to 0.920.
The KENO model has a slightly negative bias ( AK= -0.001) deduced from NAA's benchmarking calculations.
Comparing the uppet bound 95% confidence interval result of the KENO nominal case and the PDQ runs, and based on the KEN 0 benchmarking results, a calculational bias of 0.026 in AK is applied to the adjusted PDQ Xeff to provide the transport to diffusion theory correction.
(c) Summary of Results Keff (PDQ) adjusted (froc (a))
0.915 Transport to diffusion theory correction (from (b))
0.026 Final Keff 0.941 Design Limit, Keff 0.950 Calculational Margin AK 0.009 The final Keff value (0.941) includes all the design specification tolerances, model bias, and the 95% confidence interval from the KENO calculations.
6.2.3.6 Abnormal Conditions (a) Assembly Drop Accident No adverse reactivity effect is expected from dropping a fuel assembly on top of a fully loaded storagc rack during fuel han6 ting because of the large_ water thickness (NIO inches) existing between the top of the assemblies already inside the cavities and the' dropped assembly resting on top of the rack. Moreover, the basic calculational-model assumes an infinite fuel length in the l
axial direction.
13
The con.equence; of dropping an assembly outsiJe the rack and parallel to an assembly located in the outermost row of the storage array were analyzed. The assembly dropped during handling is assumed to lodge parallel to an assembly in an outer cavity with no Boral slab separating the two assemblies.
The dimensions of the rack units and the spacing is the same as the refernce case, although the dropped fuel is unchanneled to nccoit the closest contact with the rack.
Due to the reflective boundary conditions used, the array is repeated in three directions. The fourth boundary uses a zero flux cond! tic.n.
The analysis produced an increase in reactivity less than +0.5 AK in a 4 x 4 array with 8 inches of pool water in the outer cavity.
(b) Interspersed Moderation 3
Water densities from.15 to.75 gm/cm were analyzed. The resultant Keff's are:
-4 group PDQ -7 model fraction of tull Keff water density
.15
.5691
.25-
.6276
.35
.6845
.50
.7556
.75
.8392 1.00
.8931
'- KENO IV Model
.15
.5301+.0023
.25
'59587.0029 1.00
.9124[.0043 These results indicate that the limiting moderation condition exists when water 3
density is 1.0 gm/cm. Therefore misting is
.nou a significant problem.
- 6.2.4
- High Density Spent' Fuel Rack Design e
14 r
5 m
w e
t
Spent fuel storage racks provide a place in the fuel pool for storing new and spent fuel. The high density spent fuel racks are of a bolted anodized aluminum construction containing a neutron-absorbing redium of natural boron carbide (B4C).in an aluminum matrix core clad with 1100 series aluminum. This neutron absorber is marketed under the trade name of Boral. The core has a thickness of 80 mils and a minimum B-10 density of 0.0232g/cm. The Boral is sealed within two concentric square aluminum tubes hereinafter called
" poison cans".
Each rack consists of six basic components:
1) top grid casting 2) bottom grid casting 3) poison can assemblies 4) side plates 5)-
corner angle clips 6) adjustable foot assemblies The material dimensional characteristics of the rack are:
Boral. slab
(.125 x 5.25 x 152)in a) sheath thickness 0.0225 in b) sheath material Aluminum c)
Core material BC 35%
4 aluminum 65%
2 d)
.B C a real density 0.185 g/cm t
e)
W/o B in B C 70-77%
4 f) a/o B-10 in B 19.75 + 0.30%
2 g).
B-10 a real density 0.0232 gm/cm Inner!can a) each side 6.156 in. inside, square
-b) thickness 0.125 in c)
Material.
Aluminum-Outer can a)
'each side.
7.093 in..outside, square b) thickness 0.125 in.
c)- material Aluminum 15-
-fuel assembly pitch 6 625 in. +.25 The construction of the rad assures that all adjacent
. storage cavities are separated by a Boral slab.
Programmed and Remote Systems Corporation (PAR), the rack manufacturer, assures that correct Boral locations and quantities were present in accordance with the design and procurement documents through a rigorous quality assurance program evaluated and approved by Bechtel.
Bechtel's quality assurance program through auditing, surveillance, and complete shop inspection activities at PAR's facilities, verified implementation of PAR's QA program. Also, during fabrication PP&L performed additional independent verification of PAR's QA program. Documentation related to the high density spent fuel storage racks is located at the Susquehanna site.
Periodic verification of Boron effectiveness will be accomplished through an inservice inspection program utilizing test coupons. The effectiveness of the spent fuel storage rack is determined through periodic weighing of test specimens. The test specimens are contained in the test coupons.
Each component is anodized separately. The top and bottom grids are machined to maintain nominal fuel spacing of 6.625" inches center'to center within the rack and a spacing of 9.375" inches between centers or cavities in adjacent racks. Poison cans nest in pockets which-a're cast in every other cavity opening of the grids. This arrangement ensures that no structural loads will be imposed on the poison cans. The poison cans consist of two concentric square tubes with four Boral plates located in the annular gap. The Boral is positioned so it overlaps the fuel pellet stack length in the fuel assemblies by 1 inch at the' top and at the bottom. The outer can.is. formed into the inner can at each end and seal. welded to isolate.the Boral from the SFP. water. Each.can is pressure and vacuum leak tested. The grid structures are bolted and riveted together during fabrication by four corner angles and four side shear panels. Leveling screws are' located at
.the rack corners toJallow adjustment for variations in
_ pool floor level of up to +.75 inch. To maintain a flat, uniform contact area, the bearing ' pad at the 16
bottom of each leveling screw is free to pivot. The bottoms of the racks are at least 7 inches above the floor to allow for coolant flow under the racks.
Provisions are made for coolant flow in the corner cavities between the foot assembly and the bottom of the casting. These are top entry racks designed to maintain the spent fuel in boron-carbide / anodized aluminum poison compartments that preclude the possibility of criticality under normal and abnormal conditions. Lateral support is provided along the length of the storage cells in a manner that minimizes the application of lateral and bending lo:d: to the fuel assembly during handling and starr.ge.
The weight of the fuel asembly is supported in a chamber hole by the bottom casting.
The location of the spent fuel storage facility within the station complex is shown in Section 1.2 of the SSES FSAR. The racks are connected to wall embedments on the pool walls and shown in FSAR Figure 9.1-2b.
Each pool has 24 racks for a storage capacity of 2840 fuel bundle / assemblies plus 10 tultipurpose cavities for storage of control rods, control rod guide tubes, and l
defective fuel containers.
The rack arrangement is designed to prevent accidental insertion of fuel bundle / assemblies between adjacent racks. The design of the rack modules and their construction requirements preclude the possibility that the racks may be put together ao that adjacent fuel assemblies will not have a boral plate between them.
The racks are designed such that alternating storage cells contain poicon. The orientation of the modules when installed in the pool, is such that the
" checkerboard" pattern will be maintained across module boundaries. The rack modules are braced to the spent fuel pit walls, such that movement are braced to the spent fuel pit walls, such that movement (i.e.,
slippage) along the module boundaries is prevented.
6.3 Radiation Monitoring The training and experience of the Radiation Protection Officer is presented in Attachment 3.
I 6.3.1 Personnel Monitoring Personnel authorized to uncrate, survey, store, and inspect new fuel assemblies shall carry personnel
-17 i
L
conitoring devices. Records of exposure will be maintained on Form NRC-5 or equivalent.
6.3.2 New Fuel Assembly Monitoring Each new fuel assembly will oc monitored for removable alpha and beta contamination following unpackaging.
In addition, a beta gamma dose rate survey will be taken on each assembly. Records of results will be maintained in accordance with station procedures.
6.3.3 Area Monitoring Fuel unpackaging and storage areas will be routinely monitored for removable contamination levels and general radiation levels as applicable. Survey frequency, posting, and record keeping will be performed in accordance with, Health Physics procedures.
The criteria determining the minimum area survey frequency is based on the fuel handling activities occuring and personnel access to the area. Survey frequency will be at a maximum during periods of fuel receipt when personnel access is at a minimum.
When receipt and inspection work is in progress, area monitoring will be performed daily. Otherwise, monitoring will be done on a weekly basis.
'6.3.4 Instrumentation
.The following instrumenta lon, or equivalent, will be-available to support the receipt and inspection of new fuel bundle / assemblies:
Tne Model Range Use Gas Flow Eberline FC-2
'NA Removable Counter contamination survey Ioniza Victoreen 0 mR/hr-Radiation tion 470-A.
1000 R/hr level survey These instruments.will be calibrated on a quarterly basis.
>6.3.5 Emergency Actions
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The SSES Operations Shift Supervisor or designated alternate is responsible for initiation, direction, and termination of any emergency response.
6.3.6 Request for Exemotion Applicant requests exemption from the monitoring and emergency procedures requirements of 10CFR70.24. This exemption is requested because the nature of the special nuclear material, storage arrangenents, and procedural controls which Applicant proposes to employ preclude any possibility of accidental criticality during receipt, unloading, inspection and storage of new fuel assemblies.
7.0 FIRE PROTECTION Construction' of the fire protection systems pertaining to fuel receipt, inspection and storage will' be completed prior to receipt of fuel. Construction activities on the refueling floor involving a potential fire hazard or using fimnmable materials will be completed or controlled so as not to pose a threat to fuel being received, inspected or stored.
For a complete discussion of the Susquehanna SES fire protection program, see the Susquehanna Fire Protection Review Report (FPRR) previously submitted to the NRC.
Fire suppression capabilities.important to new fuel receipt and storage are identified below.
7.1 Railroad Airlock and Haistway (EL 670'-0"-818')
The railroad airlock and' hoistway have been designated as fire zone 1-2C (See FPRR Subsection 4.3.13). Fire suppression capability is supplied by automatic dry pipe sprinklers _and manually operated fire fighting equipment consisting of fire hose _ stations and portable xtinguishers.
e 7.2 Refueling Floor (EL 818')
-The refueling _ floor has been designated as fire zone 0-8A (See FPRR Subsection 4.3-32).. Jire_ suppression cappoility is supplied by manually operated hose stations equipped with straight stream no::les (See Attachment 4).
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8.0SlCURITY See the Susquehanna Physical Security plan which was submitted under separate cover to Mr. George McKorkle of the Physical Security Licensing Branch.
9.0 IDEWITY Application has been made by Pennsylvania Power 5 Light Conpany to l
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Anerican Nuclear Insurers for nuclear liability insurance in the amount of 51,000,000 fer the period from first shipment of fuel assemblies from General Electrics canufacturing facilities in Wilmington, North Carolina, until the first fuel assembly is loaded into the reactor. Proof of financial protection is being furnished to Ira P. Dinit:, NRC Indennity Specialist.
1 PENNSYLV/SIA PONER 6 LIQiT COMPANY i
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ATTACIDENT.3 TO PLA 594 ~ 1 of 3 RADIATION PROTECTION OFFICER - TRAINING AND EXPERIENCE w. FA'2: Dennis J. Trout kTFJ' EXPERIENCE: NUCLEAR k' ORE: ~ l Position: Acting Eealth Physics Supervisor Pe=sylvania Power & Light 'Co=pany Susquehanna SES Dates: July 31, 1979 to Present Location: Susquehanna Steam Electric Station - Eervick, PA Duties: Supervise Health Physics Sectied activities. Direct develop =ent of Ecalth Physics Progra=s and Procedures to assure imple=enta- ~ tion of ISAR co==it=ents and regulatory require =ents. Selection of qualified Health Physics personnel and imple=entation of re-quired training programs. Develop equip =ent and canpower budgets to ceet program require =ents. Assist in E=ergency Plan and is-ple=enting proce' dure develop =ent'. Serve as ce=ber of Corporate ALARA Review Co==ittee, and Radiation Monitoring Review Co=mittee. Position: Eealth Physics Engineer Pe=sylvania Power & Light Co=pany - Susquehanna SES 1 ~~ Dates: ' March 22, 1976 - July 31, 1979 Location: JEly 12, 1977 to July 31, 1979 Susquehanna Steam Electric Station - Berwick, PA March 22, 1979 to July 12, 1977 Susquehanna Stea= Electric Station Plant Staff Two North Ninth Street, Allentown, PA Duties: - Developed Nuclear St'ation "as low as reasonable achievable" (ALARA) Program. Assisted Health Physics Supervisor in the eval- . nation and/or direction of Health Physics Program and acted for ~ Eealth Physics Supervisor in his absence. Participated in the preparation of FSAR Section 12, developed Health Physics pro-cedures, organized the external and internal radiation exposure program, evaluated in selected Station Health Physics equip =ent, participated in various Health Physics Training Progra=s for Station personnel. Served as Chair =an of the Station ALARA ' Review Co==ittee. Directed emergency on-site and off-site =eni- .toring tea =s in respo:ise to March 28, 1979 accident at Three Mile Island. ' ) O L)E UN S'N[$ age 1 of 3 f P 1
~ ATTACHMENT 3 TO PLA 59f. 2 of 3 Positio5: Assisia~nt Engineer Nuclear I Three Mile Island (TMI) ~ Nuclear. Station - Metropolitan Edison Co=pany ~ Dates: March 1, 1974 - March'12, 1976 Location: Three Mile Island Nuclear Station - iiiddletova, PA Duties: Developed Health Physics procedures and i=ple=ented reporting require =ents. Performed release rate and dose calculations. I=plemented training progra=s for Station personnel and Rad / " Che=. Technicians.- Organised radiation exposure progra= and directed in-house TLD operations. I=ple=ented the Station Eealth Physics Progra= through first line supervision of P,ad/ Che=. Technicians. Supervised technicians and, assisted with Eealth Physics evaluations through initial fueling, startup, and nu=erous caintenance esti=ates for various aspects of the ,first refueling outage of Three Mile Island Unit I. Pro =oted In January,1976 to the position of Te+ical Analyst III. Position: Public Infor=ation Coordinator. Dates: June'11, 1973 - March 1, 1974 . Location: Three Mile. Island Nuclear Station - Middletevn, P'A ~ ~ Duties: Lectured to public interest groups c'encerning various aspects - of TMINS with e. phasis on environ = ental and Eealth Physics concerns. EDUCATION AND T?.AINING 1969-1973 Albright College, B.S. Physics 1975 Eealth Physics Certification Training / Refresher 1975 Penn. State Reactor Operations Course 1976 Respiratory Protection Progra=s and OSHA 1976 SSES Radioactive Waste Treateent Syste= 1977 GE Boiling Water Reactor Technology 1977 UNI Radiological Training for Maintenance Managers and Engineers, ~ 1978 GE Radiological Engineering Course 1979 Susquehanna SES Planc Syste=s Course. 1979 Packaging and Transportation of Radioactive Waste PROFESSIONAL SOCIETIES: Health Physics Society Delaware Valley Society for Radiation Safety 1 1 ~ Page 2 of 3 Pon 0RML ~
ATTACmENT 3 TO PI.A 594 'I 3 of 3 ~ ~ ~ IXPERIENCE WITH RADIXTION' Duration Isotope Amount Iocation of Use Type of Use ~ Hired Fision, Trace-Three Mile 2 years Reactor Health Activation, Kilocurie Island Nuclear t-. Physics Program & Corrosion Station Products Byproduct', cource & sPecial nuclear Haterial ? - .~ ), Page 3 of 3 ~ 5 A )i
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