ML20008F188
| ML20008F188 | |
| Person / Time | |
|---|---|
| Site: | 07002937 |
| Issue date: | 12/23/1980 |
| From: | Office of Nuclear Reactor Regulation, PENNSYLVANIA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20008F185 | List: |
| References | |
| 18622, PLA-647, NUDOCS 8103120463 | |
| Download: ML20008F188 (14) | |
Text
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TWO NOR TH NINT H S T R E E T, a L L E N 10W N, P A, 18101 P H O N E: (2 t S) 8 21-5151 1
December 23, 1980 t
Director Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C.
20555
'SUSQUEHANNA STEAM ELECTRIC STATION Docket #50-387 SPECIAL NUCLEAR MATERIALS LICENSE APPLICATION
- 70-29-37 NEW FUEL STORAGE AND RECEIPT ER 100450 FILE 841-06 PLA-594 This application is filed pursuant to Title 10 Code of Federal Regulations, Part 70 for authorization to receive, possess, store, inspect, and package for transport nuclear fuel bundle / assemblies.* This application supersedes that which was filed on October 16, 1980. The changes contained herein documents a r,ersonnel change in the position of Radiation Protection Officer and addresses' 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 exempt from licensing fees per 10CFR 170.11(3).
Any questions cencerning this application should be forwarded.to Mr. Thomas E. Gangloff, (215)-7/G-5543 of the Nuclear Licensing Group.
The.following information is submitted in-support of the application.
1.0 APPLICANT Pennsylvania Power & ' Light Company.
P.O.-Box-1870 Allentown,' PA 18105 Allegheny Electric Cooperative, Inc.
212 Locust Street Harrisburg, PA 17101 2.0 ADDRESS OF STORAGE SITE Unit: 1 of the 'Suse,dehanna Steam Electric Station (SSES) is located in Salem Township, Luzerne county, in east central' Pennsylvania, about five miles. northeast of Berwick, Pennsylvania.
- bund 1'e - that which is received from th.: fuel manufacturer assembly'- fuel. bundle with. channel affixed.
^
L 18622 810's1209(s;3l
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- 02/27/81l.
7 3.0 CORPORATE INFORMATION The information set forth 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.
in.ormation concerning control and owcership of the applicant is also set forth in the application fur Construction Permit and Operating License.
4.0 RADIOACTIVE MATERIAL Initial Core Fuel Assemblies Maximum square dimension of 5.47 in.
fuel l
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 (maximum) 166.97 in.
tb anel wall thickness
.080 in.
Rod array type
x 8 = 64 rods 62 fuel rods 2 water rods Fuel rod pitch 0 640 in.
Fuel rod clad thickness P.032 in.
Rod clad material Zircaloy Fuel rod outside diameter
.483 in.
Fuel pellet diameter
.410 in.
l Fuel pellet material UO2 Nominal fuel stack densisy 10.32 gm/cm Total number of fuel 764 (3 types) assemblies Maximum U-235 enrichment (weight %)
3.00 Average U-235 enrichment (weight %)
1.88 Average U-235 enrichment (weight %)
.711 1.76 2.19 per assembly type Number of assemblies per type 92 240 432 9222?l8MR
__M_
with additional lateral support near the center of gravity of the fuel assembly.
c)
The lower casting supports the weight of the fuel bundle / assembly and restricts the lateral movement; the center and top casting restricts lateral movement only of the fuel bundle / assembly.
J d)
The New Fuel Storage Vault Racks are made from aluminum. Materials used for construction are 7
"~
specified in accordance with ASTM specifications' in effect in 1970. 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 galianic 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.
c)
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 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 ft-lbs 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.
i)
The rack is designed to withstand horizon *.al combined loads up to 222,000 lbs, well in excess of expected loads.
j)
The maximum stress in the fully loaded rack in a faulted condition is 26.0 Kip. This is lower than the allowable stress.
02/27/81 7
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i k)
The rack is designed to handle nonirradiated, low emission radioactive fuel assemblies.
6.2.3 Criticality Control - Storage of New Fuel in High Density Spent Fuel Racks The design of the spent fuel pool storage racks assure that Keff will remainI O.95 for both normal and abnormal storage conditions. Normal conditions exist when the fuel is in residence in the fuel rack during dry, flooding and flooded condition.
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.
T) ensure that General Design Criteria 62 is met, the f>11owing normal and abnormal spent fuel storage conditions were analyzed.
a) normal positioning in the spent fuel storage
- array, b) fuel stored in multi purpose canisters, c) pool water temperature increases to 212 F, d) abnormal positioning in the spent fuel storage
- array, e) 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 abnormal conditions, Keff will remain 50.95.
A discussion of the analytical methodology to establish a reference case is herein presented.
Sensitivity studies have also been performed to account for manufacturing and construction tolerances for the Susquehanna SES fuel storage. A discussion and results of these studies are included.
6.2.3.1 The Principal Analytical Model ums -
that reactivity decreases continuously as temperature increases from 32 F.
(t ) Void Effect The effcet of boiling (assuming equal voids in:.ide and outside of the rack) is -tudied by varying the voids from 0% to 20% at a i
temperature of 212 F with the nominal g ometry.
The results indicate a continuous lecrease in reactivity for this range of voids.
(c)
Boral Width Redt..cron The nominal width of the Boral slab is 5.25".
The change in reactivity due to the -0.01 inch minimum tolerance on this width i.=
K=0.001.
(d)
Channel Effect The rack must accommodate both channeled and unchanneled fuel. Studies reveal that the chaaneled 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 A k from channeled to unchanneled fuel is 0.002 in the reference 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,4C and aluminum 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.080" 2
thickness with 0.0232 g/cm B-10 loading was calculated to be 0.969.
The vendor of the Boral slabs, Brooks and Perkins, made neutron attenuation studies for Boral slabs identical to those furnished on the Susquehanna project. They have determined that using an attenuation factor of 0.963 minimug assures a B-10 areal density of 0.0233 gm/cm minimum.
02/27/81 11
_..~
r NAI, using calculational techniques, utilizing a B-10 loading of 0.0232 gm/cm 2 l-producing an attenuation factor of 0.963
)
yields a Boral core thickness of 0.055" 0.003
)
inches.
i NAl's analysis shows the difference between an attenuation factor of 0.969 and 0.963 reveals about-0.003 AK for the Susquehanna spent fuel storage racks, which is accounted 1
for in the final reactivity table.
Therefore, the Boral acceptance criteria is 2
j.
cither a minimum B-10 loading of 0.0232 gm/cm or a minimum neutron attenuation factor of 0.963.
(f) Change in Pitch (pitch sensitivity))
4 The reactivity effect of mechanical tolerance changes caused by structural, fabrication, installation, and seismic onditions can be conservatively determined oy 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 i
reduenion in the range from 6.875 inch pitch b
to 6.375 inch pitch. Based on a total dimensional and positional tolerance of 0.250", a 0.014 AK bias is included in the final reactivity summary.
1-(g) Off-Center Positioning
~
- The reactivity et.ect of off-center p
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 4
f(a) Diffusion Theory Results The results of the criticality analysis of the = storage rack for Susquehanna Spent Fuel l
^
are summarized below:
PDQ Results:
Keff Reference Case ~
0.893~
l Dimensional and' Positional Tolerance, gK 0.014 Boron Width Effect, Ag o, col _
Temperature Effect, AK 0.004 p
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Variation or Boral Accept.
0.003 Void Ef fect, AK Negative Adjusted K for Susquehanna SES 0.915 (b) Monte Carlo Results and the Calculational Bias The KENO-IV verification calculatio: for the 4
nominal case yields a Keff value of 0.912+0.004 with 95% confidence interval ranging frcs 0.904 to 0.920.
The KENO model has a slightly negative bias (; K= -0.001) deduced from NAI's benchmarking calculations.
Comparing the upper bound 95% confidence interval result of the KENO nominal case and the PDQ runs, and based on the KENO benchmarking results, a calculational bias of 0.026 in LK is applied to the adjusted PDQ Keff to provide the transport to diffusion theory correction.
(c) Summary of Results Keff (PDQ) adjusted (from (a))
0.915 Transport to diffusion theory correction (from (b))
0.026 Final Keff for Susquehanna SES 0.941 Design Limit, Keff 0.950 Calculational Margin aK 0.009 for Susquehanna SF" The final Kef f.a.tur (0.941) includes all the design speci~
- v. ion tolerances, codel 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 storage rack during fuel handling because of.the large water thickness ( e10 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 assumas an infinite fuel length in the axial direction.
02/27/81 13
_ _ _ _ _ _ _ _ _ _ _ _ _ _. ~
r Spent fuel storage racks provide a place in the fuel pool for storing new and spent fuel.
The high density spent fuel tacks are of a bolted and riveted anodized l
aluminum construction containing a neutron-absorbing medium i 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.
Boral slabs are manufactured by Brooks and Perkins under a proprietary qualified process. This process assures a uniform minimum B-10 density of 0.0233gm/cm2 in the Boral slabs utilized in the construction of the Susquehanna Racks.
Benchmark measurements of those slabs yield a neutron attenuation factor of 0.963 minimum.
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. The construction of the rack assures that all adjacent storage cavities are separated by a Boral slab. The Boral is sealed within two concentric square aluminum tubes referred to as poison cans.
Each rack consists of six basic components:
1) top grid casting 2) bottom grid casting 3) poison cans 4) side plates 5) corner angle clips 6) adjustable foot assemblies 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 of cavities in adjacent racks.
Poison cans nest in pockets which are 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 tube is crimped at both ends and seal welded to the inner tuba to isolate the Boral from 02/27/81 15
_ =.
s 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 to allow adjustment for variations in pool floor level of up to +.75 inch.
To maintain a flat, uniform contact area, the bearing pad at the 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 flaw in the corner cavities between the foot assenbly 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 cornal and abnormal conditions.
Lateral support is provided along the length of the storage cells in a manner that minimizes the application of lateral and bend _ng loads to the fuel assembly during handling and storage. The weight of the fuel asembly is supported by the chamfered hole in the bottom casting.
The material dimensional characteristics of the rack are:
Boral Slab (sandwich) a.
thickness caximum 0.100 in.
einicum 0.075 in.*
b.
width 5.25 in. nominal c.
length 152 in. nominal d.
sheath material 1100 series, aluminum e.
core material B C + aluminum 4
f.
boron isotcpic content natural (19.75% B-10 nominal) 2 h.
B-10 areal density 0.0233 gm/co, minimum i.
Neutron attenuation factor 0.963 minimum
- Minimum dimension is based on review of existing quality verification documents.
Inner can a.
. width 6.156 in. 1D nominal b.
length 157.8 in.
nominal c.
thickness 0.125 in.
nominal d.
material 5052 series aluminum
4 Outer Can a.
width 7.093 in. OD nominal b.
length 153.7 in, nominal c.
thickness 0.125 in.
nominal d.
material 5052 series aluminum Poison Can Length 157.8 in.
nominal Fuel a.ssembly pitch 6.625 in. +.25 Dechtel' 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 the spent fuel storage racks will be accomplished through an inservice insnection program utilizing test specimens.
The test sacimens will be periodically weighed.
'ine rack arrangement is designed to prevent accidental insertion of fuel bundle /assembliec between adjacent racks.
The design of the rack modules and their construction requirements preclude the possibility that the racks may be put together so that adjacent fuel assemblies will not have a boral plate between them.
The racks are designed such that alternating storage cells contain poison.
The orientation of the modules when installed is the pool, is such that the
" checkerboard" pattern will be maintained across nodule boundaries.
The rack modules are braced to the spent fuel pool walla, such that loads are transferred to the spent fuel pool walls and movement (i.e., slippage) along the module boundaries is prevented.
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 c=bedments 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 multipurpose cavities for storage (eg. control rods, control rod guide tubes).
02/27/81 17
_. _,. _. _. ~ _... _ _ _ _ ~ ~
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.. _. - -... -.. ~. -
i 6.3 Padiation Mon: oring 4
l The training and experie ce of the Radiation Protection Officer i
is presented in Attachm.it 3.
i t
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6.3.1 Personnel Monitoring i
j Personnel aitthorized to uncrate, survey, store, and i
inspect new fuel assemb.ies shall carry personnel i
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02/27/81:
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Rev. 19, 1/81.
SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 -
FINAL SAFETY ANALYSIS REPORT REFERENCE CASE FUEL-STORAGE POISON CAN 02/27/81 FIGURE 9.14b
m.. _
Attachnent 3 to PLA-594 Page 1 of 2 RADIATION PROTECTION OFFICER - TRAINING AND EXPERIENCE Nan _:
Michael R. Buring t
Education and Yraining 1970 Ohio State University - B.S.-Zoology Work Experience 1962-1967 U.S. Navy - Enlisted Nuclear Plant Operator, Engineering Lab Technician, Prototype Instructor Duties:
Mechanical Operator / Instructor at Naval Nuclear Power Plant Prototype, Health Physics and Water Chemistry Control, both Primary and Secondary.
1976-1970 Batelle Memorial Institute - Safety Technician Duties:
Inspection and Auditing of various research projects in progress, for compliance with established procedures and regulatory requirements, 1970-1973 Virginia Electric and Power Company - Surrey Power Station Health Physicist Duties:
Assist station Health Physicist in routine and special projects, personnel dosimetry, radwaste, radiochemistry, procedure writing, radiological environmental monitoring.
"+s.
19P3-19J9 Metropolitan Edison Company - Corporate Radiation Safety w..
Duties:
Technic ~di~ kgort, of TMI station personnel in Health Physics, personnel dosimetiy*;W.vas,te, procedure writing and review, radiological environmental mh.dtoving, etc.
Supervised personnel dosimetry group during and after acciucr?-
"~~.n._
1979-Present Pennsylvania Power & Light Company - Environmental Grop 'ddyrc: 12 ?r,..,
u Nuclear Duties:
Supervise the implementation of radiological and non-radiological environmental monitoring programs.
Licenses and Certificates None 02/27/81
- s ou D
-g Experience With Radiation - Michael R. Buring Isotope Amount Location Duration of Use Type of Use Mixed Fission, Trace ~
U.S. Navy 12 Years Power and Naval Activiation, & Corrosion Kil.ucuric Surrey Nucicar Sta.
Reactor llealth Products - Byproduct, Three Mile Island Physics Source and Special Nuclear Station Physics Program Nuclear Material 2N
~e O"E "w
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17.
ND
r PP&L TWO NO RTH NINTH $?REET, ALLENTO WN, PA.18101 PHONE (215) 770 5151 I
NORMAN W. CURTG l
%ce President-Engmeenn2 & Constrt.ction-Nuc' ear l
7745381 February 27, 1981 l
Director Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, D.C.
20555 SUSQUEHANNA STEAM ELECTRIC STATION SPECIAL NUCLEAR MATERIALS LICENSE APPLICATION-AMENDMENT NEW FUEL STORAGE AND RECEIPT l
ER 100450 FILE 841-06 DOCKET #50-387 PLA-647
- 70-29-37 The attached pages amend PP&L's December 23, 1980 application for a Sini l
Fuel Storage and Receipt License (PLA-594). The changes contained herein document a personnel change in the position of the Radiation Protection l
Officer and addresses concerns expressed by your Norm Ketzlach.
The December 23, 1981 transmittal may be updated by following the collating instructions below:
Remove from PLA-594 Insert into PLA-594 pages 1, 2, 7, 8, 11, 12, 13, pages 1, 2, 7, 8, 11, 12, 13, i
15, 16, 17 15, 16, 17, 17a Attacament 3 This application is exempt from licensing fees per 10CFR170.ll(3).
Any questions regarding this application should be forwarded to Mr. Thomas E.
Gangloff, (215) 770-5543 of the Nuclear Licensing Group.
Very truly yours,
-i i
\\
N. W. Curtis Vice President-Engineering & Construction-Nuclear Attachments TEG:meb
.m P E N N $ Y LV A NI A P O W E R 8.
LIG H T CO MP ANY
4 PPat TWO NORTH NIN T H STR E E T, A L L E N TOWN, P A. 18101 PH O N E : (215) 8 21 5151 i
l December 23, 1980 Director j
Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Coannission Washington, D.C.
20555 SUSQUEHMNA STEAM ELECTRIC STATION Docket #50-387 SPECIAL NUCLEAR MATERIALS LICENSE APPLICATION
- 70-29-37 NEW FUEL STORAGE AND RECEIPT ER 100450 FILE 841-06 PLA-594 This application is filed pursuant to Title 10 Code of Federal Regulations, Part 70 for authorization to receive, possess, store, inspect, and package for transport nuclear fuel bundle / assemblies.* This applicqt La supersedes that which was filed on October 16, 1980. The changes contained herein documents a personnel change in the position of Radiation Protection Officer and adshesses 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 exempt from licensing fees per 10CFR 170.11(3).
Any questions concerning this application should be forwarded to Mr. Thomas E. Gangloff, (215)-770-5543 of the Nuclear Licensing Group.
The following information is submitted in support of the application.
1.0 APPLICANT Pennsylvania Power & Light Company P.O. Box 1870 y
l Allentown, PA 18105 l
l Allegheny Electric Cooperative, Inc.
212 Locust Street
(
Harrisburg, PA 17101
(
l-2.0 ADDRESS OF STORACE SITE Unit 1 of the Susquehanna.:. cam Electric Station (SSES) is located in Salem Township, Luzerne Q unty, in cast central Pennsylvania, about five miles northeast of Berwick, Pennsylvania.
r i
- bundle - that which is received from the fuel manufacturer assembly - fuel bundle with channel affixed.
P E N N 5 Y L V A NI A POWER & LIGHT COMPANY.
02/27/81 L
.-m
i 1
3.0 CORPORATE INFORMATION The information set forth 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 af the applicant is also set forth in the application for Const: action Permit and Operating License.
i 4.0 RADI0 ACTIVE MATERIAL Initial Core Fuel Assemblies Maximum square dimension of 5.47 in.
fuel l
Active fuel length 150 in.
Overall fuel bundle / assembly length 176.16 in.
s Maximum square dimension of fuel channel 5.45 in.
Overall channel 3.ength (maximum) 166.97 in.
Channel wall thickness
.080 in.
Rod areay type 8 x 8 = 64 rods 62 fuel rods 2 water rods i
Fuel rod pitch 0.640 in.
Fuel rod clad thickness 0.032 in.
l Rod clad material Zircaloy Fuel rod outside diameter
.483 in.
Fuel pellet 4iameter
.410 in.
l l
Fuel pellot material UO2 Nominal fuel stack density 10.32 gm/cm Total number of fuel 764 (3 types) l assemblics I
Maximum U-235 enrichment (weight %)
3.00 l
l Average U-23! enrichment (weight %)
1.88 l
Average U-235 en-ichment (weight %)
.711 1.76 2.19 per assembly type Number of assemblies per type 92 240 432 3
\\
.02/27/81 2
i with additional lateral support near the center of gravity of the fuel assembly, c)
The lower casting supports the weight of the fuel bundle / assembly and restricts the lateral movement; the center and top casting restricts lateral movement only of the fuel bundle / assembly.
d)
The New Fuel Storage Vault Racks are made from aluminum. Materials used for construction are specified in accordance with ASTM specifications in effect in 1970. 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 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 ft-lbs 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
'orce 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.
i)
The rack is designed to withstand horizontal combined loads up to 222,000 lbs, well in excess of expected loads.
j)
The maximum stress in the fully loaded rack in a faulted condition is 26.0 Kip. This is lower than the allowable stress.
02/27/81 7
4 k)
The rack is designed to handle nonirradiated, low emission radioactive fuel assemblies.
6.2.3 Criticality Control - Storage of New Fuel in High Density Spent Fuel Racks The design of the spent fuel pool storage racks assure that Keff will remain 10.95 for both normal and abnormal storage conditions.
Normal conditions exist when the fuel is in residence in the fuel rack during dry, flooding and flooded condition.
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 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) fuel stored in multi-purpose canisters, c) pool water temperature increases to 212*F, I
d) abnormal positioning in the spent fuel storage
- array, e) 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 abnormal conditions, Keff will remain 50.95..
A discussion of the analytical methodology to establish a reference case is herein presented.
Sensitivity studies have also been performed to account for manufacturing and construction tolerances for the Susquehanna SES fuel storage. A discussion and results of these studies are included.
6.2.3.1 The Principal Analytical Model (IR/87/81 8
/
that reactivity decreases continuously as temperature increases from 32*F.
(b) Void Effect The effect of boiling (assuming equal 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 5.25".
The change in reactivity due to the -0.03 inch minimum tolerance on this width is A K=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 unchannelcd fuel is dropped. The decrease in a k from channeled to unchanneled fuel is 0.002 in the reference 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,4C and aluminum powder An adjustment must be made to take into a. :ount directly the self-shielding effect ue to the random distribution of B C 4
l grains of finite sizes in the mixture.
Based on a homogeneous mixture, the neutron attenuation factor for a Boral slab of.080" i
2 thickness with 0.0232 g/cm B-10 loading was calculated to be 0.969.
The vendor of the l
l Boral slabs, Brooks and Perkins, made neutron l
attenuation studies for Boral slabs identical l
to those furnished on the Susquehanna proj ect.
They have determined that using an attenuation factor of 0.963 minimug assures a B-10 areal density of 0.0233 gm/cm minimum.
l t
02/27/81 11
J NAI, using calculational techniques, 2
utilizing a B-10 loading of 0.0232 gm/cm producing an attenuation factor of 0.963 yields a Boral core thickness of 0.055"!0.003 inches.
NAI's analysis shows the difference between an attenvation factor of 0.969 and 0.963 reveals about-0.003 AK for the Susquehanna spent fuel storage racks, which is accounted for in the final reactivity table.
Therefore, the Boral acceptance criteria is either a minimum B-10 loading of 0.0232 gm/cm' or a minimum neutron attenuation factor of 0.965.
(f) Change in Pitch (pitch sensitivity))
The reactivity effect of mechanical tolerance changes caused by structural, fabrication, installation, and seismic conditions can be conservatively determined by the p: tch 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.
Based on a total dimensional and positional tolerance of 0.250", a 0.014 AK bias is included in the final reactivity summary.
(g) Off-Center Positioning 5
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 storage rack for Susquehanna Spent Fuel l
are summarized below:
PDQ Results:
Keff Reference Case 0.893 l
Dimensional and Positional Tolerance,AK 0.014 Boron Width Effect,AK 0.001 Temperature Effect, AK 0.004 02/27/R1 12
]
s Variation on Boral Accept.
0.003 Void Ef fect, AK Negative Adjusted K for Susquehanna SES 0.915 (b) Monte Carlo Results and the Calculational Bias The KENO-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 05 K= -0.001) deduced from NAI's benchmarking calculations.
Comparing the upper bound 95% confidence interval result of the KENO nominal case and the PDQ runs, and based on the KENO benchmarking results, a calculational bias of 0.026 in AK is applied to the adjusted PDQ Keff to provide the transport to diffusion theory correction.
(c) Summary of Results Keff (PDQ) adjusted (from (a))
0.915 Transport to diffusion theory correction (from (b))
0.026 Final Keff for Susquehanna SES 0.941 Design Limit, Keff 0.950 Calculational Margin aK 0.009 for Susquehanna SES 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 storage rack during fuel handling because of the large water thickness ( ~10 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 axial direction.
02/27/81 13
o 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 and riveted anodized l
aluminum construction containing a neutron-absorbing medium 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.
Boral slabs are manufactured by Brooks and Perkins under a proprietarv qualified process. This process assures a uniform ninimum B-10 density of 9.0233gm/ cad in the Boral slabs utilized in the construction of the Susquehanna Racks.
Benchmark measurements of those slabs yield a neutron attenuation factor of 0.963 minimum.
Programmed and Remote Systems Corporation par), the rack manufacturer, assures that correct Boral locations and quantities we re present in accordance with the design and procurement documents through a rigorous quality assurance program evaluated and approved by Bechtel. The construction of the rack assures that all adjacent storage cavities are separated by a Boral slab. The Boral is sealed within two concentric square aluminum tubes referred to as poison cans.
Each rack consists of six basic components:
1) top grid casting 2) bottom grid casting 3) poison cans
'4) side plates 5) corner angle clips 6) adjustable foot assemblies 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 of cavities in adjacent racks.
Poison cans nest in pockets wnich are 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 tube is crimped at both ends and seal welded to the inner tube to isolate the Boral from 02/27/81 15
I 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 to allow adjustment for variations in pool floor level of up to +.75 inch. To maintain a flat, uniform contact area, the bearing pad at the 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 2anner that minimizes the application of lateral and bending loads to the fuel assembly during handling and storage. The weight of the fuel asembly is supported by the chamfered hole in the bottom casting.
The material dimensional characteristics of the rack are:
Boral Slab (sandwich) a.
thickness maximum 0.100 in.
~
minimum 0.075 in.*
b.
width 5.25 in. nominal c.
length 152 in. nominal d.
sheath material 1100 series, aluminum e.
core material B C + aluminum 4
f.
boron isotopic content natural (19.75% B-10 " "I"*1) 2 h.
B-10 areal density 0.0233 gm/cm, l
minimum 1.
Neutron attenuation factor 0.963 minirum
- Minimum dimension is based on review of existing quality i
verification documents.
l Inner Can a.
width 6.156 in. ID nominal l
l b.
length 157.8 in.
l nominal c.
thickness 0.125 in.
nominal d.
material 5052 series l
aluminum t
I l
l 02/27/81 16 c
o Outer Can a.
width 7.093 ir,. OD nominal b.
length 153.7 in.
nominal c.
thickness 0.125 in.
nominal d.
material 5052 series aluminum g
Poison Can Length 157.8 in.
nominal I
i Fuel assembly pitch 6.625 in. +.25 Bechtel' quality assurance program through auditing, surveillance, and complete shop inspection activities
. at par's facilities, verified implementatian 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 the spent fuel storage racks
[
will be accomplished through an inservice inspection program utilizing test specimens. The test specimens will be periodically weighed.
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 l
the racks may be put together so that adjacent fuel assemblies will not have a boral plate between them.
The racks are designed such that alternating storage cells contain poison. The orientation of the modules shen installed in the pool, is such that the
" checkerboard" pattern vill be maintained across module boundaries. The rack modules are braced to the spent fuel pool walls, such that loads are transferred to the spent fuel pool walls and movement (i.e., slippage) along the module boundaries is prevented.
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 conne :ted 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 multipurpose cavities for storage (eg. control rods, control rod guide tubes).
0 02/27/81 17
6.3 Radiation Monitoring The training and experience of the Radiation Protection Officer is presented in Attachment 3.
6.3.1 Personnel Monitoring Personnel authorized to uncrate, survey, store, and inspect new fuel assemblies shall carry personnel l
I 1
I t
e 02/27/81 17a L
Attechment 2 to PLA-594 3
6.156" E
E-5215" r
'i
~
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- !a channes -
l!il r:l!.
0.120"
'h g
Il iI 0.080"
- ~
ll At can v
p; l
'N VolO i
O.047" II l"l wease holes
! l i-N I
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water gap J.
- T*
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0.3505" d
we=
c ll 988 il I..;i 0.0475" i
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l Rev. 19, 1/81 SUSOUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 l
FINAL SAFETY ANALYSIS REPORT REFERENCE CASE FUEL STORAGE POISON CAN 02/27/81 l
l FIGURE 9.14b
~
Attechment 3 to PLA-594 Paga 1 of 2 RADIATION PROTECTION OFFICEA - TRAINING AND EXPERIENCE Name:
Michael R. Buring Education and Training
~
1970 Ohio State University - B.S.-Zoology Work Experience 1962-1967 U.S. Navy - Enlisted Nuclear i'lant Operator, Engineering Lab Technician, Prototype Instructor Duties:
Mechanical Operator / Instructor at Naval Nuclear Power Plant Prototype, Health Physics and Water Chemistry Control, both Primary and Secondary.
1976-1970 Bate 11e Memorial Institute - Safety Technician Duties:
Inspection and Auditing of various research projects in progress, for compliance with established procedures and regulatory requirements.
1970-1973 Virginia Electric and Power Company - Surrey Power Station Health Physicist Duties:
Assist station Health Physicist in routine and special projects, personnel dosimetry, radwaste, radiochemistry, procedure writing, radiological environmental monitoring.
1973-1979 Metropolitan Edison Company - Corporate Radiation Safety Duties:
Technical Support of TMI station personnel in Health Physics, personnel dosimetry, radwaste, procedure writing and review, radiological environmental monitoring, etc.
Supervised personnel dosimetry group during and after accident.
1979-Present Pennsylvania Power & Light Company - Environmental Group Supervisor-
!bclear Duties:
Supervise the implementation of radiological and non-radiological environmental monitoring programs.
Licenses and Certificates None 02/27/81 5,.6+
M.~.
e S.,
o Ny Experience With Radiation - Michael R. Buring Isotope Amount Location Duration of Use Type of Use Mixed Fission, Trace -
U.S. Navy 12 Years Power and Naval Activiation, & Corrosion Kilocurie Surrey Nuclear Sta.
Reactor llealth Products - Byproduct, Three Mile Island Physics Source and Special Nuclear Station Physics Program Nucicar Material IN
%0 Pa"
~
kb O
m g
e
. CC m
-M U