ML21005A012
| ML21005A012 | |
| Person / Time | |
|---|---|
| Site: | 07109356 |
| Issue date: | 12/28/2020 |
| From: | NAC International |
| To: | Office of Nuclear Material Safety and Safeguards |
| References | |
| ED20200119 | |
| Download: ML21005A012 (161) | |
Text
- December 2020 Revision 20C
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I NAC
~NTE RNAT~O NAl Atlanta Corporate Headquarters* 3930 East Jones Brldge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.naclntl.com
Enclosure I to ED20200119 Page I of2 Enclosure 1 Proposed CoC Changes
- No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion Initial Submittal MAGNATRAN SAR, Revision 20C
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FI CATE NUMBER b. REVISION NUMBER c. OOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAG E PAGES 9356 3 71-9356 USA/9356/B(U)F-96 1 OF 42
- 2. PREAMBLE
- a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the appli cable safety standards set forth in Title 10, Code of Federal Regulations, Part 71 , "Packaging and Tra nsportation of Radioactive Material."
- b. This certificate does not rel ieve the consig nor from compliance with any requirement of the regulations of the U.S. Department of Transportation or other applicable regulatory agencies , including the government of any country through or into which the package will be transported .
- 3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION
- a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFI CATION OF REPORT OR APPLICATION NAG-International NAC International , Inc., application dated 3930 East Jones Bridge Road July 1, 2019, as supplemented .
Norcross, GA 30092
- 4. CONDI TIONS Th is certificate 1s conditional upon fulfilling the requir 0 G i PaG, as aP.plicable, and the conditions specified below.
,. .::,V"' l..""I (a) Packaging ~ >-o
- (1)
(2)
Model No.:
Description :
The MAGNA
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C, purpose system for the storage and tr po MAGNASTOR transportable s ~.... ,..,.,,.,., g includes the package body, upper an y consists of the inner and outer shells, le --*-* _ ::::*:.,, _.,.... ....'--- g and solid neutron shield .
Leakage testing of the cask containment seals assures that the containment does not leak. The TSC is credited for moderator exclusion , thus serving the 10 CFR 71 .55(c) function of being a special design feature that prevents a single packing error from permitting leakage into the fissile material region . Regardless of credit applied to the TSC confinement boundary to prevent water in leakage, the containment function is retained by the transQort cask body.
The packaging body is a cylinder with multiwall construction consisting of inner and outer stainless steel shells separated by a lead gamma radiation shielding . The inner and outer stainless steel shells are 1.75 and 2.25 inches thick, respectively. The lead gamma shield is 3 .2 inches thick. Welded above the inner and outer steel shells is the upper forging . The upper forging is 7.2 inches thick where it attaches to the inner and outer shells.
The bottom of the package body consists of the bottom inner forging, the bottom outer forging and the bottom plate . The bottom inner forging is cup shaped and welded to the inner shell and the bottom forging . The ring-shaped bottom outer forging is welded to the outer shell and to the bottom plate . The bottom plate is welded onto the outer ring. The bottom inner forg ing is 5 inches thick and the bottom plate is 8.65 inches thick for a total of 13.65 inches of stainless steel shielding through the bottom .
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUM BER b. REVISION NUM BER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U) F-96 2 OF 42 5.(a)(2) Description (continued)
The package lid is a 7.75-inch-thick stainless steel disk used to close the package . The lid is attached to the top forging by forty-eight, 2-8 UN-2A socket head cap screws. The socket head cap screws screw into the tapped holes in the upper forging. The package lid is sea led by two concentric O-rings, as is the coverplate for the lid port, using inner metallic and outer ethylene propylene diene monomer (EPDM) O-rings. The MAGNATRAN package contains a lid port that is closed by a bolted Type 304/304L stainless steel coverplate with dual O-rings.
There are four stainless steel coverplate bolts. The lid port provides access to the port opening and the quick-disconnect fitting for backfilling and sampling the cavity gas during loading and unloading .
The neutron shield is comprised of NS-4-FR encased in stain less steel enclosures . The neutron shield material and its enclosure have two thicknesses , 5.8 inches and 6.4 inches, and is attached onto the outer shell alo~ he I <:,the active fuel region around the circumference of the package cavity. ~~ (J{.
Two diametrically ing trunnions are bolted t'1t side of the top forging to lift the transport packa ransport, the lifting trunnion oved and replaced with
- trunnion plugs package to pe longitudinal tie approximately direction.
. ~
nions are locat
......,....-..:;=;.u..,,.-,,,
- -;r,;_,....,,,g 0
II near the bottom of the
- tions and to provide ions are located
. sk rotates in the proper A cavity space Ii u o canister and to minimize excessive longitudi is sized to accommodate the long TSC. 0 ~~~Mi~~~, ,
The MAGNATR onsisting of a combination of redwood and b e impact limiters are bolted over each end of the pa event. The impact Ii retaining rods and nuts. ** * -
re attached to the lid ar,i cting The TSC is constructed of a stainless steel cyli ndrical shell, bottom-end plate, closure lid, age during a package drop plate via 16 tapped holes for closure ring , and redundant port covers. The TSC confines the fuel basket structure and the spent fuel or the Greater-Than-Class C (GTCC) waste basket liner and GTCC waste. The TSC cylindrical shell is dual certified 304/304L stainless steel with a 72-inch diameter and is 1/2 inch thick and either 191 .8 or 184.8 inches long , depending on the contents. The bottom end plate is welded onto the lower end of the TSC shell and is 2.75 inches th ick. The closure lid is 9 inches thick and is either a solid stain less steel closure lid or stainless steel/carbon steel closure lid .
The closure lid is welded onto the upper end of the TSC shell. The dual port covers provide a dual-welded closure system for the vent and drain ports . The GTCC TSC is simila r in design
- and construction to the TSC's for spent fuel , but instead of a basket, it contains a GTCC waste liner.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 3 OF 42 5.(a)(2) Description (continued)
The PWR fuel basket design is an arrangement of 21 square, stainless steel fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The 21 tubes develop 37 positions within the basket for the PWR spent fuel.
Each PWR basket fuel tube has a nominal 8.86-inch square opening. Each developed cell fuel position has a nominal 8.76-inch square opening . The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube. Each neutron absorber sheet is covered by a thin stainless steel sheet to protect the neutron absorber during fuel loading and to keep it in position . The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube.
The PWR damaged fuel basket is designed to store up to four damaged fuel cans in the damaged fuel basket assembly in the short TSC. The damaged fuel basket assembly has a capacity of up to 37 undama
- which includes the four damaged fuel can locations. A dam h of the four damaged fuel can basket locations. The arra same as in the standard fuel basket, but the de he four corner ents is modified with additional structural suppo larged positio uel can at the outermost
- corners of the f opening. A da fuel can locatio~
Similar to the develop 87 b R
tI amaged fuel
'-~--... n undamage
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ominal 9.80-inch square be loaded in a damaged sket fuel tube has a nominal 5.86-inch squ nominal 5.77-inch square opening . The B inder configuration by side and corner sup ~,~1:t:t1~ s. The fuel tubes support an enclosed neutro A .,m,,.......... ~ ...........,,~, uel tube for criticality control.
Each neutron a el to protect the neutron absorber during on absorber and stainless steel cover are secured across the width and along the length of the fuel tube.
- ~
The damaged fuel can confines the fuel material within the can to minimize the potential for dispersal of the fuel material into the TSC cavity. The side plates that form the upper end of the damaged fuel can are 0.15-in thick and the tube body walls are 0.048-in thick (18-gage sheet) .
The damaged fuel can lid plate and bottom thicknesses total 11 /16 inches and the lid overall height is 2.32 inches. The damaged fuel can bottom plate thickness is 5/8 (0 .625) inch . The damaged fuel can is designed in two lengths: an overall length of 166.9 inches with a nominal cavity length of 164.0 inches; or an overall length of 171.8 inches with a nominal cavity length of 169.0 inches (shorter fuel assemblies may be accommodated with a fuel assembly spacer to limit axial movement) . For the shorter damaged fuel can , a spacer is used in the damaged fuel basket assembly or alternatively fixed to the damaged fuel can bottom plate to provide an overall height of 171 .5 inches. The damaged fuel can (DFC) lid and bottom include screened drain holes.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71 -9356 USA/9356/B(U)F-96 4 OF 42 5.(a)(2) Description (Continued)
The stainless steel GTCC waste basket liner is designed to hold GTCC waste and dimensionally fit in a TSC . The GTCC waste basket liner is 173 inches long with a 1-inch-thick bottom plate welded onto it. The GTCC liner stainless steel shell is 2 inches thick for structural and gamma shield functions , and has lifting lugs welded on the inside diameter of the shell. The liner design also includes an outer ring and a middle support under the bottom plate and drain holes in the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC includes a sump location in the bottom plate and the closure lid includes a drain tube assembly to enable draining and drying of the loaded TSC .
The package has approximate dimensions and weight as follows:
Cavity diameter 72 inches Cavity length G 193 inches Package ches lmpa r nches Pac limiters 21
- 5.(a)(3)
The maximum gro s Drawings iters
- lbs 0
The MAGNAT rdance with NAC drawings:
71160-500, Rev 6P sk, MAGNATRAN 71160-501 , Rev. @ RAN 71160-502 , Rev. 6B 71160-504, Rev. 2 ~~ NATRAN 71160-505 , Rev. 6P ... V~ Lid A AGNATRAN 71160-506 , Rev . 1 .. , Cask Cavity Spacer RAN 71160-511 , Rev. 1 Rersonn
- Stiipping Configuration , Transport Cask, MA(;NA 71160-512 , Rev. 1 Nameplate, MAGNATRAN 71160-530, Rev . 1 Misc. Details, Impact Limiter, MAGNATRAN 71160-531 , Rev . 2P Impact Limiter, Transport Cask, MAGNATRAN 71160-551 , Rev . 10P Fuel Tube Assembly , MAGNASTOR - 37 PWR 71160-559 , Rev . 0 Lifting Trunnion , Transport Cask, MAGNA TRAN 71160-571 , Rev. 1OP Details, Neutron Absorber, Retainer, MAGNASTOR - 37 PWR 71160-572 , Rev . 9P Details, Neutron Absorber, Retainer, MAGNASTOR - 87 BWR 71160-574 , Rev. 6 Basket Support Weldments, MAGNASTOR - 37 PWR 71160-575 , Rev. 11 P Basket Assembly , MAGNASTOR - 37 PWR 71160-581 , Rev. 5 Shell Weldment, TSC , MAGNASTOR 71160-584 , Rev. 8 Details, TSC , MAGNASTOR 71160-585 , Rev . 13 TSC Assembly, MAGNASTOR 71160-591 , Rev . 8P Fuel Tube Assembly , MAGNASTOR - 87 BWR
NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKE T NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 5 OF 42 5.(a)(3) Drawings (Continued) 71160-598, Rev. 7P Basket Support Weldments, MAGNASTOR - 87 BWR 71160-599 , Rev. SP Basket Assembly , MAGNASTOR - 87 BWR 71160-600 , Rev. 5P Basket Assembly, MAGNASTOR - 82 BWR 71160-601, Rev . 0 Damaged Fuel Can (DFC), Assembly, MAGNASTOR 71160-602 , Rev. 1 Damaged Fuel Can (DFC) , Details, MAGNASTOR 71160-620, Rev. 1P Top Fuel Spacer, MAGNASTOR 71160-671 , Rev. 2P Details , Neutron Absorber, Retainer, For DF [Damaged Fuel]
Corner Weldment, MAGNASTOR - 37 PWR 71160-673 , Rev. 1 Damaged Fuel Can (DFC), Spacer, MAGNASTOR 71160-674, Rev. 4P DF Corner Weldment, MAGNASTOR 71160-675, Rev. 3P DF Basket Assembly, 37 Assembly PWR , MAGNASTOR 71160-681, Rev . 1 DF, Shell Weldment, TSC, MAGNASTOR 71160-684, Rev. 2 Detai s, Di flos ce Lid, MAGNASTOR 71160-685 , Rev. 8 P. SQ; Assem !'.JY, , GNASTOR 71160-711 , Rev . 1 V\,,6TCC Waste Baske ~1t1e MAGNASTOR 71160-781, Rev. 1 ~ Shell Weldment, GTCC TS~ MAGNASTOR 71160-785, Rev. 4~ GTCC TSC, Assembly, MAG , STOR Contents
( 1) 4i"'"~
Type and Form of Mater
~~" '- ,
(i) Undamag <C WR ' ~~~ 0 Undamaged p ..;:.,.;u* ...,.........., r U , IJl"lC>I\ mbl imum.
assembly avera ,B;;~~m .nRi.!.~ ** ...,:;, pent nuclear fuel that does not have any visi ....,.,.,.,......,,.,,,,* curs in the reactor, assemblies that issing rods that are replaced by solid stainles me equal to or greater than the original rods lies th efects that adversely affect radiological and/or en afety r result i ed fuel rod lengths in excess of 60 inches and that can be hand e0 6t
- e As. Undamaged PWR fuel is loaded into the short TSC, except for Combustion En (CE) 16x16 fuel assemblies , which may be loaded into either length TSC.
The fuel assemblies consist of uran ium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy instrument and guide tubes. Empty fuel rod positions are to be filled with a solid filler rod or a solid neutron absorber rod . PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC . Prior to irradiation, the fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 1. In addition , the PWR fuel must meet the fuel class assembly specifications listed in Table 2 .
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 6 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)
The burnup credit loading curve in Table 3, must be used for the 37 assembly loading profile.
WE15x15 fuel may use the burnup credit loading curve in Table 4, with the 33 , 35 or 36 assembly loading scheme provided the required cell locations for that profile shown in Figure 1 are left empty, at a minimum. Fuel assembly burn up, minimum initial average enrichment 1 , and cool tim e requirements are provided in Table 9 and Table 11 , for PWR baskets with Type 2 neutron absorbers and Table 12 and Table 14 for baskets with Type 1 neutron absorbers .
Burnup credit curves are only applicable to systems not crediting moderator exclusion . Initial enrichment up to 5 wt. % 235 U, with no burnup requirement, is permitted when crediting moderator exclusion .
Unirradiated fuel and unenriched fuel are not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod m sed as replacement rod to return a fuel assembly to an undamaged condition. \,~ G (Jl.
s emblies may contain nonfu oware . Fuel assemblies with an instrument tube tie pair shall be loaded with fuel in /or top spacers to ensure
- proper spacing a required when supported. Th components, in 1v1 u Partial-length installed . No
/ro h
f ort of the fuel assembly. Fuel in
"'=!~~
fuel tube basket e: aus
/or top spacers are not p nozzle is adequately mplete assembly or as individual ds or attial-length rods/rodlets.
ovide It ide tube plug devises are or cobalt-60 activity requirements b ~~~!if~ f ardware must meet the additional coo e o orbers), and Table 15 (for Type 1 neutron Hafnium absorb jf~~~~~~~~f , tinghouse (WE) assemblies and may have a ma """'""""' ve a minimum cool time of 16 years. Fuel a iated nonfuel solid filler fuel replacement rods . Activated
- cement rods are limited to 5 steel rods per assembly, 1 assembly per ba teel rod exposure of 32 .5 GWd/MTU . Fuel assemblies with activated stai e cooled for either a minimum of 21 years or the loading table minimum cool time (as adjusted for additional cool times for nonfuel hardware, as applicable) plus 1 year, whichever is greater.
Fuel assemblies loaded with in-core instrument thimbles must meet the additional cool time requirements in Table 5 or Table 15, as appropriate, for BPRAs or GTPDs, whichever is bounding, for Westinghouse and B&W fuel types and for Reactor control components (RCCs) for CE fuel types. The additional cool time requirements for assemblies with nonfuel hardware are added to any additional cool time requirements due to damaged fuel also being loaded in the same TSC. Reactor control components (RCCs) are restricted to fuel storage locations 11, 1
Assembly average fuel enrichment is the enrichment value determined by averaging the entire fuel region (U02) of an individual assembly, including axial blankets, if present.
235 U wt% enrichment over the
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. OOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 7 OF 42 5.(b)(1)(i) Contents - Type and Form of Material (continued) 12, 13, 18, 19, 20, 25, 26 and 27 in Figure 1. Only one neutron source assembly (NSA) is permitted to be loaded in a TSC in fuel storage locations 11 , 12, 13, 18, 19, 20, 25, 26 or 27, as shown on Figure 1.
NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers , in addition to containing burnable poison rodlets and/or thimble plug rodlets. NSAs, guide tube thimble plug devices (GTPDs) , and burnable poison rod assemblies (BPRAs) are not authorized for CE fuel assembl ies . In addition , the following un-irradiated nonfuel hardware may be loaded with the fuel assemblies:
stainless steel rods inserted to displace guide tube "dashpot" water, instrument tube tie components , and guide tube anchors or similar devices. Axial power shaping rods are not allowed contents .
Under-burned Westinghouse sem~ mblies with a maximum enrichment greater than that
- p credi oa ing may be loaded provided that a rod cluster control as CA) is inserted in the as the enrichment is equal to or less 2
than 4.05 wt.% he assembly burnup is great or equal to 12,000 MWd/MTU.
- When loading stainless steel having an exp absorber sheet without an RC assembl *
~~
el, the RCCAs mu
\c~-, d with 80% ~ ~ ' I l g1h Ag-In-Cd RCCAs comprised of a
~.,,,..;..;..;;_;,;_;,;,,;,,;.;,
5% Cd absorber pellets and
. T e basket must include ny assemblies loaded r the applicable or RCAA insertion are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt. %
23 5
- burnu re uirement is ermitted when creditin moderator exclusion.
I assemblies or nonfuel
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICAT E NUMBER b. REVISION NUMBER c. DOCKET NUM BER d. PACKAG E IDE NTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 8 OF 42 Table 1 - PWR Hybrid Fuel Assembly Characteristics No. of Min Min Max Max No. of Guide Max Max Hybrid Hybrid Clad Clad Pellet Active Vendor Array Fuel Tubes Pitch Load Assembly Group OD Thick. OD Length Rods (See (in.) (MTU)
Note 1 (in .) (in .) (in.) (in .)
BW BW15H1 H1 15x15 208 17 0.5680 0.4300 0.0265 0.3686 144 .0 0.4807 BW BW15H2 H2 15x15 208 17 0.5680 0.4300 0.0250 0.3735 144.0 0.4807 BW BW15H3 H3 15x15 208 17 0.5680 0.4280 0.0230 0.3742 144 .0 0.4807 BW BW15H4 H4 15x15 208 17 0.5680 0.4140 0.0220 0.3622 144.0 0.4690 BW BW17H1 H1 17x17 264 25 0.5020 0.3770 0.0220 0.3252 144.0 0.4681 CE CE14H1 H1 14x1 4 5 0.4400 0.0260 0.3805 137.0 0.4115 CE CE16H1 H1 0.0250 0.3250 150.0 0.4463 WE WE14H1 H1 0.3674 145.2 0.4144 WE WE15H1 H1 21 0.3669 144.0 0.4671 WE WE15H2 H2 21 0.3570 144.0 0.4469 WE WE17H1 H1 25 0.3232 144.0 0.4671 WE WE17H2 H2 0.3088 144.0 0.4327
(')
Notes:
- 1. Combined number of guide and instru 0
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') ~o
- ~
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAG E IDENTIFICATION NUMBER PAGE PAG ES 9356 3 71 -9356 USA/9356/B(U)F-96 9 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)
Table 2 - PWR Fuel Class Assembly Characteristics Fuel Class Characteristic 14x14 14x14 15x15 15x15 16x16 17x17 BW, SPC ,
Base Fuel Type 2 CE, SPC WE, SPC WE, SPC BW, FCF CE WE, FCF Max Initial Enrichment (wt. % 235 U) 3 5.0 5.0 5.0 5.0 5.0 5.0 Min Initial Enrichment (wt. % 235 U) 3 1.3 1.3 1.3 1.3 1.3 1.3 Number of Fuel Rods 4 176 179 204 208 236 264 Max Assembly Average Burnup 60,000 60,000 60,000 60,000 60 ,000 60 ,000 (MWd/MTU) 5 Min Cool Time (years) 4 4 4 4 4 4 Max Weight per Storage Location (lbs.) See N~te'1,_~~a NM E e erJjte 1 See Note 1 See Note 1 See Note 1 Max Decay Heat per Fuel Location (Watts) 6 te~~o't2 See Note 2 See N~ ~" :~ See Note 2 See Note 2 See Note 2 u
~
Notes:
- 1. Maximum weight per *
- 1,765 lbs (wei * ' s~ t fuel assembly, nonfuel hardware, damaged n cers) with ontents weight of 62,160 lbs for the PWR basket and 61, x* nominal assembly length is 178.3 inches for ass i .-=-;'="°* s ies in the short TSC . The maximum nominal fu h
- 2. For PWR baskets wit =-.,,,.,111. = =-cc- *.n~~;;;- r maximum heat load is 622 watts per storag io e burnup
>45 ,000 MWd/MTU) , and for ==ri::r-. y neutron absorbers the maximum heat load is 59.5 w ,,.......-~if""t".!..., ~"li:'1-ff'-.....,w r um assembly average burnup >45 ,000 MWd/Mit.' ) . uti~ef,r m the nonfuel hardware .
1/4.. "V/) ~0 s
- ~
2 Indicates assembly and/or nuclear steam supply system vendor/type referenced for fuel input data. Fuel acceptability for loading is not restricted to the indicated vendor provided that the fuel assembly meets the load limits . Abbreviations are as follows : Westinghouse (WE), Combustion Engineering (CE), Siemens Power Corporation (SPC) , Babcock and Wilcox (BW) , and Framatome Cogema Fuels (FCF) .
3 All reported enrichment values are nominal preirradiation fabrication values.
4 Assemblies may contain nonfuel hardware and/or fuel replacement rods (also referred to as filler rods) . Filler rods are considered to be a component of spent nuclear fuel assemblies and not nonfuel hardware. Filler rods may be burnable absorber rods , stainless steel rods or zirconium alloy rods .
A.5 Assembly average burnup is the burnup value determined by averaging the burn up over the entire fuel region (UO2) of
~ n individual assembly , including axial blankets, if present. All fuel with burnup >45,000 MVl/d/MTU is treated as damaged fuel and is placed into damaged fuel cans.
6 Maximum uniform heat load per storage location .
7 TSC and maximum contents shall not exceed 104,500 pounds
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 7 1 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUM BER b. RE VISION NUMBER c. DOCKET NUM BE R d. PACKAG E IDENTIFICATI ON NUMBER PAG E PAG ES 9356 3 71-9356 USA/9356/B(U)F-96 10 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)
Table 3 - Maximum Initial Enrichment- 37-Assembly Undamaged Fuel 15 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt% 23su) 109 Burnup = C4 x Burnup (GWd/MTU) + Cs Assembly Absorber Maximum Burn up 18 ~ Burnup Burnup ID (g/cm 2 ) Enrichment (GWd/MTU) < 18 (GWd/MTU) ~ 30 (GWd/MTU) > 30 (wt%)
C4 Cs C4 Cs C4 Cs BW15 1.9 0.0501 1.69 0.0693 1.65 0.0748 1.60 BW17 1.9 0.0502 1.72 0.0687 1.70 0.0742 1.66 CE14 2.1 0.0473 2.04 0.0675 2.03 0.0759 1.93 CE16 0.036 2.1 0.0464 2.J)l_ 0.0657 2.06 0.0733 1.99 0.04..9~ ~ ~os: ('. I / ():.0,,.6 72 1
WE14 2.2 2.21 0.0725 2.29 WE15 1.9 It"' '!I . .,,.-, ,
,.0;-Q494 1.74 -vA 0.0.2§9 ~ 1.72 0.0742 1.67 WE17 1.9
""~ .- 0.0494 1.71 0.0685/:_ 1.68 0.0687
- J,... 1.59 0.0749 1.61 BW15 1.8~ - 0.0507 1.61 0.0745 1.48 BW17 U la ~ :::::_..,Q.0503 1.66 0 .0§_89 q? ?",:1,.63 0.0733 1.59 CE14 'j.J 1't ~~0~~ 1.95
,. _,9!066!( --1.97 0.0738 1.90 CE16 0.030 12.J ")1b1tZ,0'\ cit;9~ .... (J.J!f~ 9 {1~99 0.0727 1.90 WE14 WE15 WE17 BW15
..:..2,,1 A"'~ 9 t:";{ 9 It... ~ -0~Qft
\l.1,8 ~
- 11.vi,;. ,
~
.... ft OJ)S°'O 2 } 72.~01 ~ bf ~q80
,~
~d.{![675 1.64 ~ !~ K§);:061}5/ I 1!_'5 8 ~ ~0.Q6~}U,,
2 .'1:0 t 66 1~58
,..1.52 0.0728 0.0747 0.0737 0.0754 2.19 1.54 1.53 1.41 1
BW17 ,11.:&- :ov0;lo3 .d...
"" c~ tt{Q_fl"fl,-,M t~t62 &:- 0. 0'6 ~3'
, ' ~~~59 0.0748 1.47 CE14 CE16 WE14 0.027
?.'(\
2.'t"..-
2 .1\I'_.,,,
f',
0.07(6~
0.0499 ~1, 1
-~ mta~-
1 af 92 .l-11 0".'0966
~~~l3~ / 010657 ~ ~ -92 e;_t 92
--rcY. 0667 ,.:_ ..- 2.10 0.0729 0.0747 0.0743 1.87 1.75 2.07 WE15 1.9 "'fl i- "0.0503 '1.63 o.o6cii~ 1.60 0.0749 1.46
- ~
WE17 1.9 * , o.0~ 97 1.60 .. 0.068*3' 1.54 0.0749 1.41
- .;-,r
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
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Table 4 - Maximum Initial Enrichment- Undamaged Fuel Configuration WE15x15 Optional Configurations - 20 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt % 235 U)
Number of 10s Burnup = C4 x Burnup (GWd/MTU) + Cs Assemblies Absorber Maximum Burnup 18 !S Burnup Burnup Loaded (g/cm 2 ) Enrichment (GWd/MTU) < 18 (GWd/MTU) s; 30 (GWd/MTU) > 30 (wt.% 235 U) C4 Cs C4 Cs C4 Cs 36 2.0 0.0497 1.93 0.0681 1.99 0.0747 2.00 35 0.036 2.1 0.0507 1.97 0.0673 2.08 0.0730 2.12 33 2.2 0.0504 2.12 0.0664 2.29 0.0745 2.32 36 2.0 0.0494 1.87 0.0687 1.90 0.0737 1.93 35 0.030 2.0 0.04,99 r, CJ .,9] " 0.0688 1.97 0.0740 1.99 33 2.1 "
0 ToJ!ef1 J,,}, .*
11
- 2.*oo '-" ('JQ.0686 i.15 0.0724 2.29 36 2.0 ...._ \. 11 ~. 0501 1.83 0..'0677;.. 1.87 0.0741 1.84 35 0.027 2.0..."V 0.0494 1.89 o.o6isr, 1.94 0.0735 1.96 33 2:'1"--...,,._ 0.0492 2.03 0.067_1.,'""'; t\ 2.12 0.0730 2.21
~~
_{3 ~~ho. J,.
~
~"
"Table 5 - i ~~~£a~5.~l f ss'~
/,.
-i
~
Cool Tim e" ReEioired to L;.oa~~~nf~el' Hardware '23kW/Pac ~ e)
- ~ ~*
Core l,: k-,.,. <'@ ~~~ dditiorfa~ Gool~ fme (,Years)! ,
(Assembly) ~ rJ3PRA~i!:ttt(~~!~
CE14 WE14 n
~,
Aili Gll?D ~ ~ ,1R0C I I-I 0.11 ~
~.l)-
1/fl Y#0.3 f ti[,!
,.;;;;11
~
~
NSA 1 .1 WE15 *,)) 1:3_ )'~.- 1-0~rn:,,., -~ .'t) 6.9 0 1.3 BW15 CE16
~A0,1 .
y'A..
l J etfl,o.~~
"1q/)_+~ .*
t . Iii 0.3._'-'
0t'4:~.'
0.2 WE17 tt'.4') 0.2 .;_5-y 1.4 BW17 0.1" "4
- 0.2 . ~
- 0_3 0.2
...,- r
~*~
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
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Table 6 - Nonfuel Hardware Max Exposure and Required Cool Times Years)
Maximum Minimum Cool Time (Years)
Exposure WE 14x14 WE 15x15 B&W 15x15 WE 17x17 B&W 17x17 Hardware (GWd/MTU)
BPRA 70 8.0 8.0 8.0 8.0 8.0 GTPD 180 8.0 8.0 8.0 8.0 8.0 Note: 1. Specified minimum cool times for BPRAs are independent of the requ ired minimum cool times for the fuel assembly containing the BPRA
- 2. Specified minimum cool times for GTPDs are independent of the required minimum cool times for the fuel assembly containing the GTPD.
- 3. The maximum exposure and minimum cooling time limits for NSAs without absorber rods are the same as those for GTPDs while the a imum exposure and minimum cooling time limits for NSAs with absorber rod are t sam as hose for BPRAs.
4 . Only GTPDs that do n~.t! ncl de absorber, or poison , rods or water displacement rods are allowed contents . .;:,'-' )-.__
- Hardware BPRA GTPD 894 .0 93 .3 B&W 17x17 27 .0 107 .8 Note: 1. Interpolation between exposure - GQOI time li mits is not allowed .
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Figure 1 - Undamaged Fuel Basket Loading Profile!!
(2)
(6) (7)
TSC ALIGNMENT MARK
~
~o 33 assembly loading: remove 19, 18, 20 , 12 Note: The 33 , 35 and 36-Assembly patterns also apply to the damaged fuel basket.
8 A short loaded 33 , 35 or 36 assembly loading profile may still use the burnup cred it curve in Table 4 provided that, at a minimum , the required cell locations for that profile shown in Figure 1 are left empty.
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Table 9-Loading Table for PWR Fuel - 23 kW/Package 1 Minimum Initial Assembly Average Burnup ::; 30 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14x14 14x14 15x15 15x15 16x16 17x17 17x17 2.1 ::; E < 2.3 5.7 5.8 6.7 6.9 6.3 6.8 6.8 2.3 ::; E < 2.5 5.7 5.8 6.6 6.9 6.2 6.7 6.7 2.5 ::; E < 2.7 5.6 5.7 6.6 6.8 6.1 6.6 6.6 2.7 ::; E < 2.9 5.5 5.6 6.5 6.7 6.0 6.6 6.6 2.9 ::; E < 3.1 5.6 5.6 6.4 6.7 6.0 6.5 6.5 3.1 ::; E < 3.3 5.4 5.6 6.4 6.6 6.0 6.5 6.5 3.3 ::; E < 3.5 5.4 5.9 6.4 6.4 3.5 :c; E<3.7 5.9 6.4 6.4 3.7 ::; E < 3.9 6.3 6.3 3.9 ::; E < 4.1 5.4 6.2 6.3 6.3 4.1 ::; E <4.3 6.1 6.3 6.3
- 4.3 ::; E < 4.5 4.5 ::; E < 4.7 4.7 ::; E < 4.9 E 2 4.9 Minimum lniti Assembly Av 6.1 6.1 6.2 6.2 6.1 6.1 d/MTU 6.2 6.2 6.1 6.1 Enrichment WE B&W wt% 23s u E 17x17 17x17 2.1 ::; E < 2.3 2.3 ::; E < 2.5 9.1 9.1 2.5 s E < 2.7 9.0 9.0 2.7 s E < 2.9 8.9 8.8 2.9 s E < 3.1 .8 8.8 8.7 3.1 :c; E < 3.3 7.7 8.6 8.6 3.3 ::; E < 3.5 6.7 7.7 8.6 8.6 3.5 s E < 3.7 6.7 7.6 8.5 8.5 3.7 ::; E < 3.9 6.6 6.8 8.9 7.5 8.4 8.4 3.9 s E < 4.1 6.5 6.7 8.2 8.8 7.5 8.4 8.4 4.1 ::; E < 4.3 6.5 6.7 8.2 8.7 7.4 8.3 8.3 4.3 s E < 4.5 6.4 6.6 8.1 8.7 7.4 8.2 8.2 4.5 s E<4.7 6.4 6.6 8.1 8.6 7.3 8.2 8.2 4.7 s E < 4.9 6.4 6.6 8.0 8.6 7.3 8.1 8.1 E 2 4.9 6.3 6.5 8.0 8.5 7.2 8.1 8.1
- 1. '-' means not allowed
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Table 9-Loading Table for PWR Fuel - 23 kW/Package 1 (continued)
Minimum Initial 35 < Assembly Average Burnup s 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14x14 14x14 15x15 15x15 16x16 17x17 17x17 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 9.7 11.9 13.5 14.7 11.6 13.7 13.7 2.7 s E < 2.9 9.5 10.1 13.3 14.4 11 .5 13.4 13.4 2.9 s E < 3.1 9.3 9.8 13.1 14.1 11.3 13.2 13.2 3.1 s E < 3.3 9.1 9.7 14.0 11 .1 13.0 13.0 3.3 s E < 3.5 9.0 10.9 12.8 12.8 3.5 s E < 3.7 8.9 10.8 12.7 12.6 3.7 s E < 3.9 12.5 12.5 3.9 s E<4.1 12.3 12.3 4.1 s E < 4.3 12.2 12.2 4.3 s E < 4.5 12.1 12.1 4.5 s E < 4.7 12.0 12.0 4.7 s E < 4.9 12.0 11.9 E ~4 .9 11.9 11.9 Minimum lniti d/MTU Assembly Av Enrichment WE B&W wt% 23su E 17x17 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7 s E < 2.9 20.0 20.0 2.9 s E < 3.1 19.6 19.6 3.1 s E < 3.3 19.4 19.3 3.3 s E < 3.5 16.3 19.1 19.1 3.5 s E < 3.7 12.9 16.0 18.8 18.8 3.7 s E<3.9 12.7 13.7 18.3 19.9 15.8 18.7 18.6 3.9 s E < 4.1 12.5 13.5 18.1 19.7 15.6 18.4 18.4 4.1 s E < 4.3 12.3 13.3 17.9 19.6 15.4 18.3 18.3 4.3 s E < 4.5 12.1 13.1 17.7 19.4 15.3 18.1 18.0 4.5 s E<4.7 12.0 13.0 17.6 19.2 15.2 18.0 17.9 4.7 s E<4.9 11 .9 12.8 17.4 19.0 15.0 17.7 17.8 E ~ 4.9 11 .8 12.7 17.3 19.0 14.9 17.6 17.6
- 1. '-' means not allowed
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 16 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)
Table 10-Loading Table for High Burnup PWR Fuel - 21.85 kW/Package 1 Minimum Initial 45 < Assembly Average Burnup::;; 50 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt% 23su E 14x14 14x14 15x15 15x15 17x17 17x17 2.1 ::;; E < 2.3 2.3 ::;; E < 2.5 2.5 ::;; E < 2.7 2.7 ::;; E < 2.9 29.5 2.9 ::;; E < 3.1 21.7 25 .2 28.4 30.3 28 .9 28.8 3.1 ::;; E < 3.3 21 .3 22 .9 28.2 30.1 28.7 28.6 3.3 :::; E < 3.5 21 .1 29.8 28.4 28.4 3.5 :::; E < 3.7 (. 29.7 28.1 28.1 3.7:c; E<3.9 294 28.0 28.0 3.9 :::; E < 4.1 21.4 27.3 29. 27.8 27.7 4.1 :::; E < 4.3 0.0 21 .2 27 .1 29. 27.5 4.3 :::; E < 4.5 21 .0 26.9 28. 27.4 4.5:::; E < 4.7 0.7 26 .7 8 27.2 4 .7 ::;; E < 4.9 27.0 E :::::4.9 26.9 Minimum lni d/MTU Assembly A Enrich men B&W wt% 23su 17x17 2.1 ::;; E < 2.
2.5 ::;; E < 2.7 2.7 ::;; E<2 .9 2.9 :::; E < 3.1 3.1 ::;; E < 3.3 26.8 34.9 34.9 3.3 ::;; E < 3.5 26.4 35.5 34.7 34 .6 3.5 ::;; E < 3.7 26.2 35.3 34.5 34.4 3.7 :c; E<3.9 25.9 27.9 33.1 35.1 34.4 34.2 3.9 ::;; E < 4.1 25.7 27 .6 32 .9 34.9 34.1 34 .1 4.1::;; E < 4.3 25.4 27.4 32.8 34.8 34.0 33 .9 4.3 :::; E < 4.5 25.1 27 .2 32 .5 34.6 33.9 33.8 4.5 :::; E < 4.7 25.0 26 .9 32.4 34.5 33.7 33.7 4.7 :c; E<4.9 24.7 26 .7 32 .3 34.3 33 .5 33.4 E ::::: 4 .9 24.5 26.6 32.0 34.2 33.4 33 .3
- 1. '-' means not allowed
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Table 10-Loading Table for High Burnup PWR Fuel - 21.85 kW/Package 1 (continued)
Minimum Initial 55 < Assembly Average Burn up ~ 60 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt% 23su E 14x14 14x14 15x15 15x15 17x17 17x17 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7 s E < 2.9 2.9 s E < 3.1 3.1 s E < 3.3 3.3 s E < 3.5 39 .8 39 .7 3.5 s E < 3.7 39.6 39.5 38.7 39.3 33.4 38.6 39.1 33.2 38.4 39.0 38.9 38.7 38.6 38.5
- 1. '-' mea Table 11 ----=-1::o ackage
\))
Max.
Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time MWd/MTU wt3/4 23su [Years]
10,000 1.3 4.0 15,000 1.5 4.0 20,000 1.7 4.4 25,000 1.9 5.5 30,000 2 .1 6.9
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Table 12-Loading Table for PWR Fuel - 22 kW/Package 1 Minimum Initial Assembly Average Burnup s; 30 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 :5 E < 2.3 6.0 6.1 7.1 7.4 6.6 7.2 7.2 2.3 :5 E < 2.5 5.9 6.0 7.0 7.3 6.6 7.0 7.1 2.5:5E<2 .7 5.9 6.0 7.0 7.2 6.5 7.0 7.0 2.7:5E<2.9 5.8 5.9 6.9 7.2 6.4 6.9 6.9 2.9 :5 E < 3.1 5.8 5.9 6.8 7.1 6.4 6.9 6.9 3.1:5E<3.3 5.7 6.3 6.9 6.9 3.3 :5 E < 3.5 6.8 6.8 3.5 :5 E < 3.7 6.8 6.8 3.7 :5 E < 3.9 6.7 6.7 6.7 3.9 :5 E < 4.1 6.6 6.7 4.1 :5 E < 4.3 6.6 6.7 4.3 :5 E < 4.5 6.6 6.6 4.5 :5 E < 4 6.6 4.7:5E<4 .6 6.6 E ~ 4.9 6.6 Minimum lni Assembly A Enrichme B&W wt °lo 235LJ 17X17 2.3 :5 E < 2.5 10.0 10.0 2.5 :5 E < 2.7 9.9 9.9 2.7 :5 E < 2.9 9.7 9.7 2.9 :5 E < 3.1 7.3 8.5 9.6 9.6 3.1 :5E<3.3 7.2 7.5 9.4 10.1 8.4 9.5 9.5 3.3 :5 E < 3.5 7.2 7.4 9.3 10.0 8.3 9.4 9.4 3.5 :5E<3.7 7.1 7.4 9.2 9.9 8.2 9.3 9.3 3.7 :5 E < 3.9 7.0 7.3 9.1 9.8 8.1 9.3 9.2 3.9 :5 E < 4.1 7.0 7.2 9.1 9.7 8.1 9.1 9.2 4.1 :5 E < 4.3 6.9 7.2 9.0 9.6 8.0 9.1 9.1 4.3 :5 E < 4.5 6.9 7.1 9.0 9.6 8.0 9.0 9.0 4 .5:5E<4.7 6.9 7.0 8.9 9.5 7. 9 9.0 9.0 4.7 :5 E < 4.9 6.8 7.0 8.8 9.5 7.9 9.0 9.0 E;::: 4.9 6.8 7.0 8.8 9.4 7.9 8.9 8.9
- 1. '-' means not allowed
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- 5. (b )( 1)(i) Contents - Type and Form of Material (continued)
Table 12- Loading Table for PWR Fuel - 22 kW/Package 1 {continued}
Minimum Initial 35 < Assembly Average Burnup ~ 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 $ E < 2 .3 2.3 $ E < 2 .5 2.5$E<2.7 10.7 11 .9 15.2 16.6 13.1 15.4 15.4 2.7 $ E < 2 .9 10.5 11 .2 16.2 12.9 15.2 15.1 2.9 $ E < 3.1 10.3 11.0 16.0 12.6 14.8 14.8 3.1 $ E < 3.3 10.1 10.8 12.4 14.7 14.7 3.3 $ E < 3.5 14.4 14 .5 3.5 $ E < 3.7 14.3 14.2 3.7 $ E<3.9 10.3 13.9 14.2 14.1 3.9 $ E < 4 .1 10.1 13.7 14.0 14.0 4.1 $ E < 4 . 13.6 1 13.9 4.3 $ E < 13.5 13.8 4.5 $ E < 13.6 13.6
.5 13.5 d/MTU B&W 17X17 2.3 $ E < 2 .5 2.5 $ E < 2.7 2.7 $ E < 2 .9 15.7 19.2 22.1 22. 1 2.9 $ E < 3.1 15.3 18.8 21 .8 21 .8 3.1 $ E < 3.3 15.0 16.2 22 .9 18.6 21 .5 21 .5 3.3 $ E < 3.5 14.8 15.9 20.9 22 .6 18.3 21 .3 21 .3 3.5 $ E < 3.7 14.5 15.7 20.7 22.4 18.0 21 .1 21 .0 3.7 $ E < 3.9 14.2 15.5 20.4 22 .2 17.8 20 .8 20.8 3.9 $ E < 4.1 14.0 15.3 20.2 22 .0 17.6 20 .6 20.6 4 .1 $ E < 4.3 13.9 15.0 20.0 21 .8 17.5 20 .5 20.4 4.3 $ E < 4.5 13.7 14.8 19.8 21.6 17.3 20 .3 20.3 4 .5 $ E < 4.7 13.6 14.7 19.7 21 .5 17.1 20 .1 20.1 4.7$ E <4.9 13.5 14.5 19.6 21 .3 17.0 20 .0 19.9 E ;::: 4.9 13.4 14.4 19.5 21 .2 16.9 19.8 19.9
- 1. '-' means not allowed
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
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Table 13-Loading Table for High Burnup PWR Fuel - 20.9 kW/Package 1 Minimum Initial 45 < Assembly Average Burnup :5 50 GWd/MTU Assembly Avg . r------,--- M _ i_n_im um_ C_o_o_li~n_..__T_im_e-->+--e_a_rs_,__~~--1 Enrichment CE WE WE B&W WE B&W wt 0"'o 235 U (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 :,; E < 2.3 2.3 :5 E < 2.5 2.5:5E < 2.7 2.7:,; E < 2.9 31 .0 2.9 :5 E < 3.1 23 .8 32.7 31.3 31 .2 3.1 :5 E < 3.3 31 .0 31 .0 3.3 :5 E < 3.5 30.8 30.8 3.5 :5 E < 3.7 30.5 29.8 30.3 23.6 29.6 30.1 23.4 I~~ 29.9 29.8 29 .6 29.5 29.3 TU B&W 17X17 2.5 :5 E < 2.7 2.7 :5 E < 2.9 2.9 :5 E < 3.1 3.1:5E < 3.3 28 .9 31 .7 36.1 38.1 37 .3 37.2 3.3:,; E < 3.5 28 .7 30.7 35.8 38.0 37.1 37.0 3.5:,; E < 3.7 28 .3 30.4 35.7 37.8 36 .9 36.8 3.7 :5 E < 3.9 28.1 30 .2 35.4 37.6 36 .8 36 .6 3.9 :5 E < 4 .1 27.9 29 .9 35.2 37.4 36.6 36 .5 4 .1 :5 E < 4 .3 27.6 29 .7 35.1 37 .3 36.4 36.3 4.3 :5 E < 4 .5 27.4 29.5 34.8 37.1 36.3 36.2 4.5 :5 E < 4 .7 27 .2 29 .3 34.7 37 .0 36.2 36 .1 4.7 :5 E < 4 .9 27.1 29 .1 34 .6 36.8 36.0 35.9 E 2! 4.9 26.9 28.9 34.4 36 .7 35.8 35.7
- 1. '-' means not allowed
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
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Table 13-Loading Table for High Burnup PWR Fuel - 20.9 kW/Package 1 (continued)
Minimum Initial 55 < Assembly Average Burnup !:: 60 GWd/MTU Assembly Avg. Minimum Cooling Time (years)
Enrichment CE WE WE B&W WE B&W wt% 23su (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 :s; E < 2.3 2.3 :s; E < 2.5 2.5 :s; E < 2.7 2.7 :s; E < 2 .9 2.9 :s; E < 3.1 3.1 :s; E < 3.3 3.3 :s; E < 3.5 42.2 42 .1 36.1 41 .3 41 .9 35.9 41 .1 41 .7
- 4.5 :s E 4.7 :s; E E~4 41.6 41 .5 41 .3 41 .2 41 .1 41 .0 1.
NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
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Table 14-Low Burnup PWR Fuel Loading Table - 22 kW/Package Max.
Min. Assembly Assembly Minimum Avg. Initial Avg. Cool Time Enrichment Burnup [Years]
[wt% 235 U]
[MWd/MTU]
10,000 1.3 4.0 15,000 1.5 4.0 20 ,000 1.7 4.5 25 ,000 1.9 5.7 30 ,000 2.1 7.4 f\ HE Gul.
Table 15-Add"
- Assembly Cool Time tc1Qact Nonfuel Hardware eat Load - 22kW PWR - C 1guration) 0.2
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(ii) Undamaged and damaged PWR assemblies A combination of damaged and undamaged PWR fuel assemblies in the 37 PWR damaged fuel basket, shown in Figure 2, in a short TSC . Undamaged, low burnup fuel assemblies must meet the description for PWR fuel in 5.(b)(1 )(i) . Up to four damaged fuel assemblies, high burnug fuel assemblies (maximum assembly average burnup >45,000 MWdLMB-D or fuel material that is less than, or equivalent to, one undamaged PWR fuel assembly must be placed in a damaged fuel can and must be placed in locations 4, 8, 30 and 34 in the PWR damaged fuel basket.
Undamaged , low burnup fuel may also be placed in the 4 damaged fuel locations, without the use of a damaged fuel can. Prior to irradiation, the damaged and undamaged fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 1 and meet the fuel class assembly specifications listed in Table 2.
For the 33 non-damaged fuel ti r-Fiaged fuel basket, the fuel must meet the class enrichment, po -* ia ooling time, burntiiq er:edit loading curves , and the TSC neutron absorbers density in Table 16. For th profiles up to the 33, 35 and 36 assembly loadi rn, the PWR fuel must meet th ding curves in Table 17. A
- short-loaded 3 17 provided t minimum .
fuel or hig loading cu for all location$ (burn a
,6 assembly loading profil i
, ~...:::v
- n -~'l'.I
,:u;.i.~...;..;....iii,;.ii~~c;;;...;;;.;;.;...;;...;
burnup cred it curve in Table re 1 are left empty, at a es not contain any damaged ooling time, burnup credit Tables 3 and 4 ma be used s not crediting' -_ __
exclusion). (/)
Fuel assembly ~~~:,St-ii~ f,M.~ ool time requ irements are provided in Tabl ,,,,,.,."~'-'....... rbers and Tables 12-14 for baskets with Ty aged........,..................................... fuel , all fuel assemblies i must uirements in Table 18 for the assembly type th d in the d ypes of fuel assemblies are loaded in different damaged fuel cans in a single J S gest additional fuel cooling time applies to all fuel assemblies *in the SQ. Jbe atl9itional cool time requirements in Table 18 apply to assemblies loaded in TS Bas ets with Type 1 or Type 2 neutron absorbers .
Damaged and high burn up CE 16x 16 fuel assemblies are not authorized for shipment.
The fuel assemblies consist of uranium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy instrument and guide tubes. Empty fuel rod positions for undamaged fuel assemblies are to be filled with a solid filler rod or a solid neutron absorber rod that displaces a volume equal to or greater than the original rod. PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC .
9 Assembly average fuel enrichment is the enrichment value determined by averaging the 235
- U wt% enrichment over the entire fuel region (U02) of an individual assembly , including axial blankets, if present.
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Unirradiated fuel and unenriched fuel are not authorized for loading , except that unenriched axial blankets are permitted , provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition . Damaged or high burnup fuel located in a damaged fuel can location in the damaged fuel basket must have a minimum burnup of 5 GWd/MTU , a maximum enrichment of 4.05 wt.% 235 U, and a minimum cool time of 15 years . PWR fuel assemblies loaded in a damaged fuel can must not contain nonfuel hardware with the exception of instrument tube tie components, guide tube anchors or similar devices, and steel inserts. l\pplication of moderator exclusion allows increasing the maximum initial enrichment to 5 wt. % 235 U with no burnup requ irement.
Undamaged PWR fuel assemblies may contain nonfuel hardware, while damaged PWR fuel assemblies shall not, with t ing unirradiated nonfuel hardware:
instrument tube tie com
- ilar devices, and steel inserts. The nonfuel hardware m o as individual components, individual nonfuel length rods or pa rods/rodlets . Partial-length rods/rodlets are tubes provided g g devises are installed. Fuel assemblies wit tie rod repair s with fuel inserts and/or top spacers to en ~ ~'-- nd support ~~~- . Fuel inserts and/or top spacers are n *= --= ___,,."'"" because the top nozzle is adequately su .......---.c:,_-.c. and cool time or cobalt-60 activity requir HFRAs are o ,._'lc,....,r-- ve a maximum exposure of 4.0 GWd/MT el assemblies loaded with nonfuel hardware of Table 5 (for Type 2 neutron absorbe ---..,.re* uel assemblies may contain any number of u s. Activated stainless steel fuel replacement ro assembly per basket, and a maximum steel rod of 32. lies with activated stainless steel rods must be cooled r a mini years or he loading table minimum cool time (as adjusted for additional cool ti I h rdware and the presence of damaged fuel in the TSC , as applicable) plus 1 ye is greater.
Fuel assemblies loaded with in-core instrument thimbles must meet the additional cool time requirements in Table 5 or Table 15, as appropriate, for BPRAs or GTPDs, whichever is bounding , for Westinghouse and B&W fuel types and for RCCs for CE fuel types. The additional cool time requirements for assemblies with nonfuel hardware are added to any additional cool time requirements due to damaged fuel also being loaded in the same TSC .
RCCs are restricted to fuel storage locations 11 , 12, 13, 18, 19, 20, 25, 26 and 27 in Figure 1.
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26 or 27 in Figure1. NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. NSAs, GTPDs, and BPRAs for CE fuel types are not allowed contents. In addition , the following unirradiated, nonfuel hardware may be loaded with the fuel assemblies : stainless steel rods inserted to displace guide tube "dashpot" water, instrument tube tie components, and guide tube anchors or similar devices .
Axial power shaping rods are not allowed contents.
Under-burned Westinghouse 15x15 assemblies (assemblies with a maximum enrichment greater than that dictated by the burnup credit loading curve) may be loaded provided that an RCCA is inserted in the assembly, the enrichment is equal to or less than 4.05 wt.% 235 U, and the assembly burnup is greater than or equal to 12,000 MWd/MTU . When loading under-burned fuel, the RCCAs must be full length Ag-In-Cd RCCAs comprised of stainless steel clad rods constructed with 80% Ag, 15°/o~j d % Cd absorber pellets and having an exposure equal to or less than 200,000 Wd MifU . 11 gas e must include absorber sheets with an effective 10 B areal densit ' oH l. 36 g/cm 2 . Any assen)blies loaded without an RCCA inserted must meet the burnuP, c e it loading curve for the applicabfe assembly loading profile. Burnup credit curves, and the criticality requirement for RCCA insertion, are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt. % 235 U with no burnui:2 or RCCA reguirement, is i:2ermitted when crediting moderator exclusion.
QJ
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Table 16-Maximum Initial Enrichment - 37 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt% 23 su) 108 Zero (0) = C4 x Burnup (GWd/MTU) + Cs Assembly Burnup Burnup Absorber 18 :5 Burnup 30 < Burnup 50 < Burnup ID Max. Enr. (GWd/MTU) <
(g/cm 2 (GWd/MTU) :5 30 (GWd/MTU) :5 50 (GWd/MTU)
(wt%) 18 Cs Cs Cs Cs BW15 1.6 0.0453 1.42 0.0681 1.29 0.0750 1.03 0.0750 0.736 BW17 1.6 0.0476 1.45 0.0668 1.37 0.0712 1.17 0.0712 0.891 CE14 1.9 0.0504 1.79 0.0696 1.75 0.0751 1.60 0.0751 1.60 CE16 0.036 1.9 0.0484 .._ 1t3..9 il 0&6J§l.i ~ 1.74 0.0758 1.52 0.0758 1.52 WE14 j'-,-. :~ *- " - - ""'"' j l' f Al 1.9 .Q,/ l.§1~ 1.85 0.0729 I# Jt.' *8§.._ 0.0794 1.75 0.0794 1.75 WE15 1.6 ... ~ ~'o.10482 1.43 0.0692 f'if ). 0.0738 1.08 0.0738 0. 767 WE17 1.6 ~ " 0.0439 1.45 0.0657 1.35 ~ ~ .0732 1.00 0.0732 0.700
- BW15 BW17 CE14 CE16 WE14 0.030 1.,.q .....
l-1.8
' ~,.:::~ 87 o.q~
1.31
) .71~ ~
0.0660 7gg i 1..J~
~ 64
~ 0140 o.o'"?kWI 0.896 1.37 0.0740 0.0781 0.614 1.37 WE15 WE17 BW15 1~ ~ °'~~ . rJ3.9,~~ ,~~"'s:.~-s:"- - -+-'i;i
- "-f~---19----i.s: ,._7,;i
§@ ,-7_2_5+-o_.8_5_7-+-_o._o7_2_5-+-_o._58_1--1 BW1 7 CE14 1.8 ft'J 4
,I 1.5Y-/_ 0.0474 ) 1~ fA
-.0,£)486 1.68
., If"".
0.0696 2""ti1' 1.27 _ ,'1:t.on4 1.6.'.\' ..I._. I
- !1,. -
0.0778 o.918 1.32 0.0724 0.0778 o.639 1.32 CE16 0.027 1.8 0.049~-t 1166 0.0p 60- .-., 1.6,i 0.0761 1.33 0.0761 1.33 WE14 1.8 0.0535 - :;'{ "I':l .f
'.:;,-.1" 11
.1 0.Q. 6 -14 ,._ 1.75 0.0805 1.52 0.0805 1.52 WE15 1. 5 0.0465 1.33 0.0664 1.24 0.0710 0.968 0.0710 0.685 WE17 1.5 0.0447 1.31 0.0647 1.25 0.0714 0.846 0.0714 0.564
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Table 17-Maximum Initial Enrichment - WE 15x15 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt % 235 U) 109 Zero (0) = C4 x Burnup (GWd/MTU) + Cs Number of Burn up Burnup 18 ~ Burnup 30 < Burnup 50 <
Absorber Assemblies Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Burnup (g/cm 2 )
(wt%) < 18 ~ 30 ~ 50 GWd/MTU C4 Cs C4 Cs C4 Cs C4 Cs 36 1.6 0.0483 1.53 0.0721 1.35 0.0750 1.17 0.0750 0.851 35 0.036 1.7 0.0532 1.51 0.0722 1.45 0.0778 1.14 0.0778 1.14 33 1.7 0.0524 1.60 0.0734 1.52 0.0791 1.22 0.0791 1.22 36 1.6 1.32 0.0739 1.15 0.0739 0.811 35 0.030 1.6 0.0733 1.20 0.0733 0.847 33 1.19 0.0780 1.19 36 1.02 0.0731 0.693 35 0.027 1.13 0.0738 0.775 1.09 0.0784 1.09 R Fuel Contents Max Assembly Min. lni 15x15 ;:.,_ WE 17x17 ;:.,_
Average Burnup Averag Cool Time GWd/MTU ears ears 2.5 N/A 35 0.8 0.3 3 .3 2.8 40 2 .7 1.2 0.8 2 .9 0.0 0.0 2 .7 2.6 4 .5 4.2 2 .9 2.6 2.7 2.2 45 3.1 2.5 0.7 0.1 3.3 0.0 1.0 0.0 0.0 2 .7 N/A N/A 4 .8 N/A 2 .9 3.6 2.8 3.5 2.8 50 3 .1 1.7 2.8 1.2 0.5 3 .3 0.0 1.2 0.0 0 .0 3 .1 4.2 2.9 4 .0 3.6 55 3.3 2.2 3.0 1.9 1.5 3.5 0.2 2.0 0.0 0.0 3.1 N/A N/A 5.0 N/A 3.3 4 .6 3.0 4 .9 4 .1 60 3.5 3.1 3.1 2.9 2 .1 3 .7 1.3 2.8 0.8 0.0 3.9 0.0 0 .9 0.0 0.0
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Figure 2-Damaged Fuel Basket Loading Profile (2)
(6) (7) ocation
~o
- it DFC designated locations may contain a loaded DFC or an undamaged PWR fuel assembly .
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(iii) Undamaged BWR assemblies Undamaged BWR fuel assemblies within the 87 BWR basket assembly shown in Figure 3.
Undamaged fuel is spent nuclear fuel that does not have any visible deformation other than uniform bowing that occurs in the reactor, assemblies that do not have missing rods , and assemblies with missing rods that are replaced by solid stainless steel or zirconium filler rods that displace a volume equal to or greater than the original rods and assemblies that do not contain structural defects that adversely affect radiological and/or criticality safety and/or result in unsupported fuel rod lengths in excess of 60 inches and that can be handled by normal means. BWR/2-3 assemblies are to be loaded into short TSCs, and BWR/4-6 assemblies are to be loaded into long TSCs.
The fuel assemblies consist f r *
- ellets with zirconium alloy-clad fuel rods and zirconium alloy-clad water ro s ositions must be filled with solid, unirradiated, nonfu isp ace a vo u o, or greater than , that of the fuel rod that the filler r . Prior to irradiation, th mblies must be within the dimensions and ns of the hybrid assembl able 19. In addition, the BWR fuel must meet t s assembly specificar le 20 . Fuel assembly burnup ,
minimum initial a *a ment10 , and cool
- are provided in Table 23, 24 and 25. ~ C)
Undamaged - u
- i n c;,.~~ ,
y enriG ent and the TSC neutron absorber she d .t;;;:;;ji~~ :~~~ ass y loading patterns for fuel with axial blan n "-" """"-~ "'
- ble 22. Spacers may be used to axiall io t in the TSC . Unenriched and unirradiat I i ~~4i>ii -.::.l'i.M..¥.i riched axial blankets are 1
~~,,"';wr"~'~,..,~ ~~::,~"~ ,n 1
permitted , pro t ater than 6 inches.
For a TSC that is le . fu ations shall begin with location 44, followed by locati , 45, ntinuing outward , as required, in an approximately symmetric i;ia tern as shown in igure . llowable fuel assembly locations for the 82 assembly BWR fuel asse nfiguration are shown in Figure 4 . Prior to use of the 82 assembly configuration , t eldment and upper weldments of nonfuel locations must be physically blocked (fuel storage locations 44, 32, 34 , 54 , 56 shown as in Figure 4) .
BWR fuel assemblies may be unchanneled, or channeled with zirconium-based alloy channels .
BWR fuel assemblies with stainless steel channels are not authorized.
The 82-Assembly configuration is the result of criticality constraints on maximum enrichment.
When crediting moderator exclusion this configuration is not required as full capacity (87-Assembly) is permitted at an initial enrichment U Q to 5 wt. % 235 U.
- 10 Assembly average fuel enrichment is the enrichment value determined by averaging the 235 U wt% enrichment over the entire fuel region (U02) of an individual assembly, including axial blankets, if present.
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Table 19-BWR Hybrid Fuel Assembly Characteristics Geometry 2 *3 Number Min Min Max Max Number Max Max Assembly of Partial Clad Clad Pellet Active of Fuel Pitch Loading Type Length OD Thick. OD Length Rods (inch) (MTU)
Rods 1 (inch) (inch) (inch) (inch) 87 48A 48 N/A 0.7380 0.5700 0.03600 0.4900 144.0 0.1981 87 49A 49 N/A 0.7380 0.5630 0.03200 0.4880 146.0 0.2034 87 498 49 N/A 0.7380 0.5630 0.03200 0.4910 150.0 0.2037 88 59A 59 N/A 0.6400 0.4930 0.03400 0.4160 150.0 0.1828 88 60A 60 N/A 0.6417 0.4840 0.03150 0.4110 150.0 0.1815 88 608 60 N/A 0.6400 0.483Q 0.03000 0.4140 150.0 0.1841 88 618 61 NIA o.64001\.. K oJ!i3'e.:- I ~ ,0.03000 0.4140 150.0 o.1872 88 62A 62 N/A - 0:64jt-71 0.4830 \"~ 02900 0.4160 150.0 0.1921 88 63A 63 NIA 1 ,\.d.6420 0.4840 0.0272ffi l 0.4195 150.0 0.1985 88 64A 64 N/A ..._...._ I 0.6420 0.4840 0.02725.,. 1r\ 0.4195 150.0 0.1996 88 648 4 64 N/A~ 0.6090 0.4576 0.02900 ,_.(..0.3913 150.0 0.1755
- l--_89~7_2A _-4-_7_2_-+--,;N,.,,.1A-=- "J _ '+-l~~~~ ~T!.,>{:
,i'!,2,:-0 -f--_0._43_3_0--f-----,O,.;i
.9::,--,
29biGi
,,;'!iQ7
,--t_ "1
__,,Q,.:..;.;J...,_
-" 7_4_0 -+--1_5_0_.0_+-_0_ .1_80_3--!
89 74A 74 1 _(~.J ~015~0,... 0.4240 /4cmrStl 0.3760 150.0 0.1873 89 76A 76 l N/A o~S'Z2o ~ 10;~'\pQ / ' o. QZ19o o~ iso 150.0 o.1914 810 91A 91 1 ,,8 ~ ~ ~-s'iJJ,9-,, :~ ;.39~7. :;-:~~ 2~~~/4 o.3~20 150.o o.1906 810 92A 92 1 - ,4 1 HQ)B;l,,0O~~ 'q'4'04,q ~~!(J"0.o~:eoo",l" 0.~I455 150.0 0.1946 810 96A 4 96 1 810 100A4 100
~ . u 1~~~i ~
~O Notes:
Assemblies may cont~ aiiial-length fuel rods.
2 Assembly characteristics repres t cold , unirr di ted , nominal configurations .
3 Maximum channel thickness allowetl Is 20 nfl$ (nominal) .
4 Composed of four subchannel clusters .
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Table 20-BWR Fuel Class Assembly Characteristics Fuel Class Characteristic 7x7 8x8 9x9 10x10 Base Fuel Type 11 SPC, GE SPC , GE SPC , GE SPC , GE , ABB Max Initial Enrichment (wt% 235U) 4.5 4.5 4.5 4.5 59 72 91 13 60 7413 48 61 9213 Number of Fuel Rods 76 9513, 14 49 62 79 10014 63 5412 80 Max Assembly Average Burnup (MWd/MTU) §.0,_oog 6~ 0- 000 § O-:-QQQ 60-000 Min Cool Time (years)
""" f'.
I\.R 4 Ht::f ~,. '
4 c.=i 4 4 Min Average Enrichment (wt% 235 U) 15 If"'\..,, 1.3 v ( L~ 1.3 1.3 Max Weight (lb) per Storage Locatio"-1,, "V See Note 1 See.f'R>J~ See Note 1 See Note 1 Max Decay Heat (Watts) per Fuel,~ ation 253 253 UA 253 253 Notes:
- 1. Maximum weight pe contents weight of 6
" _rag 6 lb
~i-.~:l!!:);;:r.,, er.:;,-=--=*"" space y Ieng
?
~ d channel) with a maximum 176.2 inches for BWR/4-6 assemblies and 171 J.n.fhe e ma m nominal assembly width is 5.52 inches. r-
- 2. Fuel assembly weigh I cl ::;.
- 3. Maximum initial enrich "rr,11~,., ~nt. --:i.
- 4. Water rods may occu ¥ui*r~~~-* _..._wJ" I aseybly to contain nominal number of water rods for the s
- 5. All enrichment values
- 6. Spacers may be used to
- 7. Each BWR fuel assembly ma )f * * * ~o handling .
nel s 0.120 inches thick.
11 Indicates assembly vendor/type referenced for fuel input data. Fuel acceptability for loading is not restricted to the indicated vendor/type provided that the fuel assembly meets the limits listed in Table 6.2.1-1. Table 6.2.1-2 conta ins vendor information by fuel rod array. Abbreviations are as follows : General Electric/Global Nuclear Fuels (GE),
Exxon/Advanced Nuclear Fuels/Siemens Power Corporation (SPC).
12 May be composed of four subchannel clusters.
3 Assemblies may contain partial-length fuel rods .
- 14 Composed of four subchannel clusters 15 Assembly average burn up is the burn up value determined by averaging the burnup over the entire fuel region (U02) of an ind ividual assembly, including axial blankets , if present.
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Table 21-Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel With Axial Blankets)
Max. Initial Enrichment 16 (wt % 235 U)
Fuel Absorber 17 0.027 10 8 g/cm 2 Absorber17 0.0225 10 8 g/cm 2 Absorber7 0.02 10 8 g/cm 2 Type 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Basket Basket Basket Basket Basket Basket 87 48A 4.0% 4.5% 3.7% 4 .5% 3.6% 4.4%
87 49A 3.8% 4 .5% 3.6% 4.4% 3.5% 4.3%
87 498 3.8% 4.5% 3.6% 4.4% 3.5% 4.2%
88 59A 3.9% 4.5% 3.7% 4.5% 3.6% 4 .3%
88 60A 3.8% 3.5% 4 .2%
88 608 3.8% 3.5% 4 .2%
88 618 3.8% 3.5% 4 .2%
88 62A 3.8% 3.5% 4 .1%
88 63A 3.8% ~ 4 .5% 3.6% 4 .3% -..-t\ 3.4% 4.2%
88 64A 3.5% 4.2%
88 648 3.3% 4.0%
89 72A 3.4% 4.1%
89 74A 3.7% ~ 4.3°1J ~ < 3f'.4.91,ca.J I} '1 ~.1%
1 1- ' 3.4% 4.0%
89 76A 3.5% r: ~ \~1o =. ~ 3.4% - - ~4.0°/o ,I ti; ' 3.3% 3.9%
89 79A 3.3% 4.0%
89 80A 3.5% 4.2%
810 91A 3.7% 3.5% 4.1%
810 92A 3.8% 3.5% 4 .1%
810 96A 3.7% 3.4% 4 .0%
810 100A 3.6% 3.4% 4 .0%
Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235 U for all basket/absorber combinations .
16 17 Maximum planar average.
Borated aluminum neutron absorber sheet effective areal 10 8 density.
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Table 22-Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel Without Axial Blankets)
Max. Initial Enrichment 18 (wt% 235 U)
Fuel 19 Absorber 0.027 8 i:i/cm 10 2 19 Absorber 0.0225 8 q/cm 10 2 Absorber 19 0.02 10 8 i:i/cm 2 Type 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Basket Basket Basket Basket Basket Basket B7 48A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%
B7 49A 3.7% 4.5% 3.6% 4.3% 3.4% 4.1%
B7 49B 3.7% 4.5% 3.6% 4.3% 3.5% 4.2%
B8 59A 3.8% 4.5% 3.7% 4.4% 3.5% 4.3%
B8 60A 3.7% 4.5% - 3.6% r-.... 4.3% 3.5% 4.1%
B8 60B 3.7% 4.4°/o,, ~- ~* -.3_ s Jo '-"'
0 ll £4.2% 3.4% 4.1%
B8 61B 3.7% A~"'... 3.6% ~ o"- 3.5% 4.1%
B8 62A 4.1%
B8 63A 3.7% ~ 4.4% 3.5% 4.2% ~ " 3.4% 4.1%
B8 64A 4.1%
B8 64B 4.0%
B9 72A 4.1%
B9 74A 4.0%
B9 76A 3.5% I""'.'.'.' .., \ni9!o ""' ,.,.: 3.3% .....-~4.Q0/4 / ~ 3.2% 3.8%
B9 79A 3.s3/4 u. ~ ~o/(!\ ~~ t/Di :l~wJ ~~- dlf1:74' ::=- 3.2% 3.9%
B9 80A 3.7% r\ ~gJoia~ ~ 3.ls f& ~ IId Vo ~ 3.s3/4 4.1%
B10 91A 3.7% ;\) 4AJ!,6'i...:;yr, },3~bzL.. ,. . ,. '~ _2% <,- 3.4% 4.1%
B10 92A 4.1%
B10 96A 4.0%
B10 100A 3.6% 41,% 3.4% 4~ o 3.3% 3.9%
~- 4 .
Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235 U for all baskeUabsorber combinations .
18 19 Maximum planar average.
Borated aluminum neutron absorber sheet effective areal 10 B density.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAG ES 9356 3 71 -9356 USA/9356/B(U)F-96 34 OF 42 5.(b)(1)(iii) Contents - Type and Form of Material (continued)
Figure 3-Undamaged Fuel Basket 87 Assembly Loading Profile
~ TSC ALIGNMENT MARK X = Designated NonFuel Location
NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (B-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. OOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 35 OF 42 5.(b)(1 )(iii) Contents - Type and Form of Material (continued)
Figure 4-Undamaged Fuel Basket 82 Assembly Loading Profile
/ / (1) (2) (3) ~
/ (4) (5) (6) (7) (8) (9) ( 10)
I\
(16) (17) (18) X l
(11) (12) ( 13) (14) (15)
(19) (20: (21) :22) (23) (24) (25) (26: (27) -~
- (28) (29~ (30) (31)
(39: ( 40) ( 41) (42) ~ 43)
X (33)
X X (35) (36) (37: (38)
( 45) ( 46) ( 4 7: ( 48) 1 49' \ )
(50) ( 51) (52) ~53) X (55) X (57) (58) (59: (60)
( 61) (62: (63) :64) (65) (66) (67) ( 68: (69)
\X (70) (71) (72) (73) (74) (75) (76) (77)
(78: (79) :so) (81) (82) (83) ( 84:
l/
' " (85) (86) (87)
TSC ALIGNMENT MARK X = Designated Nonfuel Location
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE
- 1 1. a. CERT FICATE NUMBER 9356 FOR RADIOACTIVE MATERIAL PACKAGES
- b. REVISION NUMBER 3
- c. DOCKET NUMBER 71-9356
- d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAG E 36 OF PAGES 42 5.(b)(1 )(iii) Contents - Type and Form of Material (continued)
Table 23-Loading Table for BWR Fuel - 22kW/Package 1 Minimum Initial Assembly Average Burn up ~30 GWd/MTU Assembly Avg. f------r-----~_M_i_n_im_u_m~C_o_o_li_n..__T_im~e-'---""'-'e_a_rs'-L--~---~---~
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235 U E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 ~ E<2 .3 6.5 12.3 5.8 13.7 5.3 13.0 13.5 2.3 ~ E < 2.5 6.3 11 .6 5.7 13.0 5.2 12.3 12.8 2.5 ~ E < 2.7 6.3 11 .0 5.7 12.3 5.1 11.7 12.2 2.7~E<2.9 6.2 10.3 5.6 11 .8 5.1 11 .1 11.6 2.9 ~ E < 3.1 6.1 9.8 5.6 11 .2 5.0 10.5 11 .1 3.1 ~ E < 3.3 6.0 9.3 5.0 10.0 10.6 3.3 ~ E < 3.5 3.5 ~ E < 3.7 6.0 6.0 8.8 ~~!R 5.4 4.9 4.9 9.6 9.1 10.0 9.6 3.7 ~ E<3 .9 5.9 5.4 9.4 -4.9 8.8 9.2 5.9 5.3 9.0 0 8.4 8.9 3.9 ~ E < 4.1 4.1 ~ E < 4.3 5.9 5.3 4. 8.0 8.5 4.3 ~ E < 4.5 5.~ 7.7 8.2 4.5 ~ E < 4.7 5(8 7.5 7.9 4.7 ~ E < 4.9 5.8 7.2 7.6 E ~ 4.9 6.9 7.4 Minimum Initial Assembly Avg. ~ - - -~ ,g=:"'-!#~-.--":::w:,;.:=p:;:r:~= ="-=-!~::;.w...:::.::.=.c:,...,....._ _ _~_ _ _---1 Enrichment BWR/4-6 BWR/4-6 wt% 23su E 9x9 10x10 2.1 ~ E < 2.3 2.3 ~ E < 2.5 15.0 15.5 2.5 ~ E < 2.7 14.1 14.6 2.7 ~ E < 2.9 8.6 13.4 13.9 2.9 ~ E < 3.1 8.5 12.7 13.2 3.1 ~ E < 3.3 8.4 11.4 12.8 6.3 12.1 12.6 3.3 ~ E < 3.5 8.3 10.8 12.2 6.2 11 .5 12.0 3.5 ~ E < 3.7 8.2 10.3 7.0 11.7 6.1 11.0 11 .5 3.7 ~ E < 3.9 8.1 9.8 6.9 11 .2 6.0 10.6 11 .0 3.9 ~ E < 4.1 8.0 9.3 6.9 10.8 6.0 10.1 10.6 4.1 ~ E < 4.3 8.0 8.9 6.9 10.4 6.0 9.7 10.1 4.3 ~ E < 4.5 8.0 8.7 6.8 10.0 6.0 9.3 9.8 4.5 ~ E < 4.7 7.9 8.6 6.8 9.6 5.9 8.9 9.4 4.7 ::; E < 4.9 7.8 8.6 6.7 9.3 5.9 8.6 9.1 E ~ 4.9 7.8 8.6 6.7 9.0 5.9 8.3 8.8
- 1. '-' means not allowed
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAG ES 9356 3 71-9356 USA/9356/B(U)F-96 37 OF 42 5 .(b)(1 )(iii) Contents - Type and Form of Material (continued)
Table 23-Loading Table for BWR Fuel - 22kW/Package 1 (continued)
Minimum Initial 35 < Assembly Average Burn up :,; 40 GWd/MTU Assembly Avg. 1---- - -~ - - -~_M_i_n_im_u_m~C_o_o_li_n..__T_im~e____..'-e_a_rs...L..-~---~-------1 Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 :,; E < 2.3 2.3 :,; E < 2.5 2.5 :,; E < 2.7 14.6 16.9 12.2 18.0 10.0 17.3 17.7 2.7 :,; E<2 .9 13.3 15.8 10.7 17.0 8.7 16.3 16.7 2.9 :,; E < 3.1 13.1 14.9 16.0 8.5 15.4 15.8 3.1 :,; E < 3.3 12.9 14.1 8.4 14.6 15.0 3.3 :,; E < 3.5 12.6 13.9 8.3 13.8 14.3 3.5 :,; E < 3.7 12.5 10.0 13.2 13.6 3.7 :,; E < 3.9 12.4 9.9 13.3 12.6 13.0 3.9 :,; E<4 .1 12.2 9.8 12.7 12.1 12.5 4.1 :,; E < 4.3 12.0 9.7 12.0 11 .9 4.3 :,; E < 4 .5 1 12.0 11 .5 4.5 :,; E < 4 .7 1 11.9 11 .2 4.7 :,; E < 4 .9 11.8 11 .2 E ~ 4 .9 11 .1 Minimum Initial Assembly Avg. 1---------.--~1"'\-H'fm'--T-'"'-'-"irl-7--'+-,-ci-/:;~-:-=-""-:rr-i:--'cd-'::-=...,-,-,t+/-,;-,---=-:-:-:-:::-:------=...,--,,,,-------I Enrichment BWR/4-6 BWR/4-6 wt% 235 U E 9x9 10x10 2.1 :,; E < 2 .3 2.3 :,; E < 2 .5 2.5 :,; E<2 .7 _
2.7 :,; E < 2 .9 22 .3 22.7 21 .5 2.9 :,; E < 3.1 19.7 14.8 20.0 19.4 3.1 :,; E < 3.3 18.9 20.5 15.4 19.1 12.3 18.8 18.2 3.3 :,; E < 3.5 18.7 20 .2 15.2 18.8 11 .9 18.6 17.4 3.5 :,; E < 3.7 18.5 20.0 15.0 18.7 11.7 18.3 17.2 3.7 :,; E < 3.9 18.2 19.9 14.7 18.5 11 .5 18.0 17.1 3.9 :,; E < 4 .1 18.1 19.6 14.6 18.2 11.4 17.9 16.9 4.1 :,; E < 4 .3 17.8 19.5 14.3 18.1 11.3 17.7 16.7 4.3 :,; E < 4 .5 17.8 19.4 14.3 18.0 11 .2 17.7 16.5 4.5 :,; E < 4 .7 17.6 19.2 14 .1 17.8 11 .1 17. 5 16.5 4.7 :,; E < 4 .9 17.4 19.0 14 .0 17. 8 11.0 17.4 16 .3 E ~ 4 .9 17.3 18.9 13.8 17 7 10.9 17.3 16.2
- 1. '-' means not allowed
NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. OOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 38 OF 42 Table 24--Loading Table for BWR Fuel - 20.9kW/Package Minimum Initial 45 < Assembly Average Burnup::;; 50 GWd/MTU Assembly Avg. Minimum CoolinQ Time (years)
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 ::;;E<2.3 - - - - - - -
2.3::;; E < 2.5 - - - - - - -
2.5::;; E < 2.7 - - - - - - -
2.7::;; E < 2.9 - - - - - - -
2.9::;; E < 3.1 29.6 31.5 27.3 30.2 24.9 30.2 29.0 3.1 ::;;E<3.3 27.8 29.6 24.7 27.9 22.2 27.6 26.3 3.3::;; E < 3.5 27.6 29.3 23.6 27.7 19.6 27.4 26.1 3.5::;; E < 3.7 27.4 29.0 23.2 27.4 19.0 27.1 25.9 3.7::;; E < 3.9 27.2 28.9 23.0 27.3 18.7 26.9 25.6 3.9::;; E < 4.1 26.9 28.6 22.8 27.0 18.5 26.7 25.5 4.1::;; E < 4.3 26.8 28.6 22.6 27.0 18.4 26.5 25.2 4.3::;; E < 4.5 26.6 28.3 22.3 26.8 18.2 26.5 25.1 4.5::;; E < 4.7 26.4 28.1 22.3 26.6 17.9 26.3 25.0 4.7::;;E<4.9 26.2 28.0 22.1 26.4 17.9 26.1 24.8 E ~ 4.9 26.0 27.8 22.0 26.4 17.8 25.9 24.7 Minimum Initial 50 < Assembly Average Burnup ::;; 55 GWd/MTU Assembly Avg. Minimum CoolinQ Time (years)
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1:s;E<2.3 - - - - - - -
2.3::;; E < 2.5 - - - - - - -
2.5::;; E < 2.7 - - - - - - -
2.7::;; E < 2.9 - - - - - - -
2.9::;; E < 3.1 - - - - - - -
3.1::;; E < 3.3 36.4 38.4 34.1 37.2 31.8 37.2 35.9 3.3::;; E < 3.5 34.0 35.8 31 .7 34.6 29.2 34.6 33.4 3.5::;; E < 3.7 33.3 35.0 29.1 33.4 26.6 33.1 31.8 3.7::;; E < 3.9 33.1 34.8 28.8 33.3 24.3 32.8 31.4 3.9::;; E < 4.1 32.9 34.6 28.6 33.1 24.0 32.7 31.4 4.1::;; E < 4.3 32.7 34.5 28.5 32.9 23.9 32.5 31 .1 4.3::;; E < 4.5 32.5 34.3 28.2 32.7 23.6 32.4 30.9 4.5 ::;; E < 4.7 32.5 34.2 28.0 32.6 23.5 32.2 30.8 4.7::;; E < 4.9 32.3 34.0 27.8 32.4 23.3 32.0 30.6 E ~ 4.9 32.1 33.9 27.6 32.4 23.2 31.8 30.5
- 1. '-' means not allowed
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FI CATE NUMBER b . REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAG E PAGES 9356 3 71-9356 USA/9356/B(U)F-96 39 OF 42 Table 24-Loading Table for BWR Fuel - 20.9kW/Package (continued}
Minimum Initial 55 < Assembly Average Burn up~ 60 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1sE<2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7 sE< 2.9 2.9 s E < 3.1 3.1 s E < 3.3 3.3 s E < 3.5 42.3 44.6 40.0 43.4 38.2 43.5 42.3 3.5 s E < 3.7 39.8 42.3 37.7 41 .0 35.7 41.1 39.8 3.7 s E < 3.9 38.3 40.0 35.3 38.7 33.3 38.7 37.5 3.9 s E < 4.1 38.1 40.2 33.7 38.3 30.9 38.0 36.5 4.1 s E < 4.3 37.8 40.0 33.6 38.2 29.0 37.8 36.4 4.3 s E < 4.5 37.8 39.8 33.4 38.2 28.9 37.8 36.4 4.5 s E < 4.7 37.7 39.6 33.2 38.0 28.7 37.7 36.2 4.7 s E < 4.9 37.6 39.5 33.1 37.9 28.4 37.5 36.0 E ~4.9 37.4 39.4 32.9 37.8 28.4 37.4 35.8
- 1. '-' means not allowed
(/)
i Table 25-1:ow
/Package Max.
Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time MWd/MTU wt3/4 23su [Years]
10,000 1.3 6.3 15,000 1.5 8.6 20 ,000 1.7 10.3 25 ,000 1.9 11.9 30 ,000 2.1 13.7
NRG FORM 618 U.S. NUCLEAR REGULA TORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 40 OF 42 5.(b)(1 )(iii) Contents - Type and Form of Material (continued)
(iv) Greater Than Class C Waste GTCC waste consisting of solid , irradiated , and contaminated hardware , provided the quantity of fissile material does not exceed a Type A quantity and does not exceed the mass limits of 10 CFR 71.15 , within a GTCC waste basket liner transported in a GTCC TSC with a welded closure lid . The specific Curie content source of the GTCC waste shall be limited to a maximum specific activity of 2.7 Ci 6 °Co/lb averaged over the GTCC waste, with a maximum localized peak specific activity of 16.1 Ci 6 °Co/lb and a total 6 °Co activity of 85 ,760 Ci at transport. The maximum allowed weight of this waste is 55 ,000 lbs.
5.(b)(2) Maximum quantity of material per package (i) For the contents described in ) undamaged PWR fuel assemblies, including nonfuel e with a maximum weight of 62,160 pounds and a max* am ecay heat limit el location not to exceed the values in Table 2. ~ )-.__
- (ii) For the conten assemblies, w damaged fuel pounds (TSC an decay heat lim1t ger f ma n
,~.-=""-
~=
5.(b)(1 )(ii) :
to 4 damag "Ji::il~~:.f~~ctl...rJ
~~=;.;.;.;.;,;;;;.,.
-=
PWR fuel
~ ~~~+/-!!!!!d i I assemblies in weight of 61 , 184 nds) and a maximum (iii) For the conte fuel assemblies, including chan pounds and a maximum dee (iv) For the contents d ith a maximum weight per package of 55 ,000 . for the GTCC waste is 5.(c)
- 1. 7 kW per packag .
Criticality Safety Index Undamaged PWR and BWR Fuel
- 0.00
-1( -i{
Damaged PWR Fuel 100.00
- 6. In addition to the requ irements of Subpart G of 10 CFR Part 71:
(a) The package must be prepared for sh ipment and operated in accordance with the Operating Procedures in Chapter 7 of the application , as supplemented .
- (b) Each packaging must be acceptance tested and maintained in accordance with the Acceptance Tests and Maintenance Program in Chapter 8 of the application , as supplemented , except that the minimum component thicknesses for the mockup in Section 8.1.6.1 and the minimum shielding effectiveness configuration for calculating the dose rates used as acceptance criteria
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 41 OF 42 for the tests in Sections 8.1.6.3 and 8.2.3 are defined by the component dimensions and tolerances in the drawings listed in Condition 5.(a)(3) .
- 7. Prior to transport by rail , the Association of American Railroads must have evaluated and approved the railcar and the system used to support and secure the package during transport.
- 8. Prior to marine or barge transport, the National Cargo Bureau, Inc., must have evaluated and approved the system used to support and secure the package to the barge or vessel, and must have certified that package stowage is in accordance with the regulations of the Commandant, United States Coast Guard.
- 9. Transport by air is not authorized.
- 10. Transport of fuel assemblies , as described in Drawing No. 71160-685 Revision 8, Assembly No. 99 , is not authorized . ~R REG
- 11. Zion TSC basket assemblies ~ numbers TSc-2Vi TSC-24, TSC-25 , and TSC-26 , are authorized to not have weld ims installed as require nse drawing 71160-575 , Note 4.
The American Society ngineers Boiler a el Code alternative in Table 2.1.4-1 , "ASME Code NATRAN ,a~~ .1.4-6 for the "Fuel Basket Assembly" the descri *
- Measures" is revised to:
"Fuel basket material =- ~~ inch, may optionally have impact tests perform .*.,u ...-....... ~ principal rolling direction of the plate, provided th ,....,,,.~~"'-- l--'1111!~~ f the measured values before comparing to --~ ~~1,/:t, lf nical Engineers Boiler and Pressure Vessel Cod ~Bi~ e~ erature (LST) of-40°F."
- 13. The package authorize byj this c --...~rl'.'-...-r-A.-, ~~~ -i.v..;,._,;:;:.,,..~ ,.,;;; use ~ er the general license provisions of 10 CFR 71. H . ~
~~ . . , ,., ., . . . ~o
- -i'
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERT FICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 42 OF 42
- 14. Expiration date: April 30, 2024 REFERENCES NAC International , Inc. , Application dated D FOR THE U.S. NUCLEAR REGULATORY COMMISSION
.Date: _ _ _ _ _ __
Enclosure 2 to ED20200119 Page 1 of3 Enclosure 2 List of SAR Changes
- No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion Initial Submittal MAGNATRAN SAR, Revision 20C
Enclosure 2 to ED20200119 Page 2 of3 0
~:
- .(
?J~~~p~erfa_g~ ~ ,~. -(
~- ~ i 't List of Changes for the MAGNATRAN SAR
!; 1; :.~ '~. .:;'
fi.: '?~*:**_,,;,.
? ,... ** - . '
{ >; ~- ~:* \** -~*- 1*~--- ~, k
/
- ,,.:1: ...
~"~*:t,~*FigurdfalJJev:,_.~* -~*~ ~ t'. -~, *:**J:,'1,**~.-:-.*,,.,*n~cription of.:Chalige,*J* ..,.,:*, -~-.;- .:,**. (:
7 ~
- ,.;, ,~* ,'.!..... ,-. ,'~ :'h ~)f .:.-*i,.,"-.:~,. ~** ;*.:,"J"r, !). ~'" 1~:*' .!' .;* ~ ~li**,.:\,'*,.1;,1,: ,*,-, ,. *-J..*. i\c:1,,~.* h- .: ,,,::.* J....;** ,.-.t Note: The List of Effective Pages and the Chapter Table of Contents, List of Figures and List of Tables have been revised accordingly to reflect the list of changes detailed below.
Chapter 1 Pages 1.3-8 thru 1.3-9 Added paragraph at the bottom of page 13-8 thru the top of page 1.3-9 where indicated.
Pages 1.3-10 thru 1.3-12 Text flow changes.
Page 1.3-21 Added text in the middle of the third paragraph in Section 1.3.2 where indicated.
Page 1.3-22 Modified text at the top of the page and near the end of the last paragraph on the page where indicated.
Page 1.3-23 Added paragraph at the top of the page.
Page 13-24 Added text near the middle of the second paragraph on the page; added new fourth paragraph where indicated.
Page 13-25 Text flow changes.
Page 1.3-26 Modified text in Items 11.b and 11.c where indicated.
Page 13-27 Text flow changes.
Page 1.3-28 Added paragraph at the end ofltem 11.p where indicated.
Page 1.3-29 Modified text in Items 12.b, 12.c, 12.d and 12.i where indicated.
Page 13-30 Text flow changes.
Page 13-31 Added text at the end ofltem 12.s where indicated.
Page 1.3-32 Added paragraph at the end ofltem 13.j where indicated.
Page 1.3-38 Deleted last bullet following Table 13-6 where indicated.
Page 13-44 Modified title of Table 1.3-12 where indicated.
Page 1.3-47 Modified Table 1.3-19 where indicated.
Page 1.3-49 Added Note below Table 1.3-21 where indicated.
Page 1.3-50 Added Note below Table 1.3-22 where indicated.
Chapter 2 Page 2.11-1 Added text to the end of Section 2.11 where indicated.
Page 2.11.1-1 Deleted text in the first paragraph of Section 2.11.1 where indicated.
Page 2.11.4-2 Added text at the end of Section 2.11.4 where indicated.
Pages 2.11.6-1 thru 2.11.6-2 Added new Section 2.11.6 where indicated.
Page 2.12.1-5 Added two references at the end of Section 2.12.1 where indicated.
Enclosure 2 to ED20200119 Page 3 of3
- ChaRter 3 No Changes ChaRter 4 Page 4-1 Added paragraph at the end of Section 4 where indicated.
ChaRter 5 Page 5-1 Modified text in the first paragraph of Section 5 where indicated.
Page 5.1-6 Replaced Figure 5.1-2 where indicated.
Page 5.1-10 Modified Table 5.1-4 where indicated.
Page 53-4 Replaced Figure 5.3-3 where indicated.
Page 5.6-26 Modified Table 5.6-4 where indicated.
Page 5.82-3 Modified text in the second paragraph of Section 5.8.2 where indicated.
Page 5.8.4-2 Modified text at the end of the first paragraph on the page in Section 5.8.4.2 where indicated.
Page 5.8.4-3 Modified the embedded table and the paragraph following it at the end of Section 5.8.4.3 where indicated.
Page 5.8.4-10 Replaced Figure 5.8-28 where indicated.
Page 5.8.4-15 Replaced Figure 5.8-31 where indicated.
Page 5.8.4-17 Replaced Figure 5.8-33 where indicated.
Page 5.8.4-20 Modified headings in Table 5.8-23 where indicated.
Pages 5.8.4-21 thru 5.8.4-22 Modified headings in Table 5 .8-24a where indicated.
Pages 5.8.4-23 thru 5.8.4-24 Added Table 5.8-24b where indicated.
Page 5.8.4-25 Modified Tables 5.8-25 and 5.8-26 where indicated.
ChaRter 6 Pages 6.1.1-1 thru 6.1.1-2 Added text in the middle of Section 6.1.1 where indicated.
Pages 6.1.2-1 thru 6.1.2-2 Modified text throughout Section 6.1.2 where indicated.
Pages 6.1.2-3 thru 6.12-6 Text flow changes.
Pages 6.4.1-1 thru 6.4.1-4 Added and modified text throughout Section 6.4.1 where indicated.
Pages 6.10.4-1 thru 6.10.4-3 Added new Section 6.10.4.
ChaRter 7 Page 7.1-6 Deleted fourth paragraph on the page Page 7.1-7 thru 7.1-16 Text flow changes.
ChaRter 8 No Changes
Enclosure 3 to ED20200119 Page 1 of2 Enclosure 3 Supporting Calculations
- for No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion Initial Submittal MAGNATRAN SAR, Revision 20C
Enclosure 3 to ED20200119
- Page 2 of2 List of Calculations:
- 1. 71160-2139 Rev. 0, PWR and BWR Fuel Assembly Fatigue Evaluation for MAGNATRAN CALCULATIONS WITHHELD IN THEIR ENTIRETY PER 10 CFR 2.390
Enclosure 4 to ED20200119 Page 1 of 1 Enclosure 4
- LOEP and SAR Page Changes No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion Initial Submittal MAGNATRAN SAR, Revision 20C
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Effective Pages Chapter 1 Page 2.6.3-1 ........................................ Revision 0 Page 1-i thru I-ii ................................. Revision 0 Page 2.6.4-1 thru 2.6.4-2 .................... Revision 0 Page I-iii ........................................ Revision 20C Page 2.6.5-1 thru 2.6.5-5 .................... Revision 0 Page 1-1 .............................................. Revision 0 Page 2.6.6-1 ........................................ Revision 0 Page I.I-I thru 1.1-8 ........................... Revision 0 Page 2.6.7-1 ........................................ Revision 0 Page 1.2-1 thru 1.2-5 ........................... Revision 0 Page 2.6.7.1-1 thru 2.6.7.1-9 .............. Revision 0 Page 1.3-1 thru 1.3-7 ........................... Revision 0 Page 2.6.7.2-1 thru 2.6.7.2-5 .............. Revision 0 Page 1.3-8 thru 1.3-12 .................... 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MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Effective Pages (cont'd)
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MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Effective Pages (cont'd)
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MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Effective Pages Chapter 8 Page 8-i thru 8-ii ................................. Revision 0 Page 8-1 .............................................. Revision 0 Page 8.1-1 thru 8.1-29 ......................... Revision 0 Page 8.2-1 thru 8.2-8 ........................... Revision 0 Page 8.3-1 thru 8.3-2 ........................... Revision 0 Page 4 of 4
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Tables Table 1.3-1 Design Characteristics of the MAGNATRAN Transport Cask and Components .................................................................................................... 1.3-16 Table 1.3-2 Design Characteristics of the Transportable Storage Canister ....................... 1.3-18 Table 1.3-3 Design Characteristics of the Fuel Basket Assemblies ................................... 1.3-19 Table 1.3-4 Design Characteristics of the Damaged Fuel Can .......................................................................................... 1.3-20 Table 1.3-5 Design Characteristics of the GTCC Waste Basket Liner and TSC ............... 1.3-20 Table 1.3-6 PWR Fuel Assembly Characteristics .............................................................. 1.3-38 Table 1.3-7 Bounding PWR Fuel Assembly Geometry for Loading Criteria .................... 1.3-39 Table 1.3-8 Maximum Initial Enrichment- 37-Assembly Undamaged Fuel Configuration - 0.036 g/cm2 1°B Absorber ..................................................... 1.3-40 Table 1.3-9 Maximum Initial enrichment- 37-Assembly Undamaged Fuel Configuration - 0.030 g/cm2 1°B Absorber ..................................................... 1.3-41 Table 1.3-10 Maximum Initial Enrichment- 37-Assembly Undamaged Fuel Configuration - 0.027 g/cm2 1°B Absorber ..................................................... 1.3-42 Table 1.3-11 Maximum Initial Enrichment - Undamaged Fuel Configuration WE 15 -
Optional Configurations .................................................................................. 1.3-43 Table 1.3-12 Maximum Initial Enrichment - PWR Damaged Fuel Configuration -
- Table 1.3-13 Table 1.3-14 Table 1.3-15 0.036 g/cm2 1°B Absorber ............................................................................... 1.3-44 Maximum Initial Enrichment - PWR Damaged Fuel Configuration -
0.030 g/cm2 1°B Absorber ............................................................................... 1.3-44 Maximum Initial Enrichment - PWR Damaged Fuel Configuration -
0.027 g/cm2 1°B Absorber ............................................................................... 1.3-45 Maximum Initial Enrichment - Damaged Fuel Configuration WE 15 -
Optional Configurations .................................................................................. 1.3-45 Table 1.3-16 Additional Fuel Assembly Cool Time Required to Load Nonfuel Hardware ......................................................................................................... 1.3-46 Table 1.3-17 Allowed BPRA Burnup and Cool Time Combinations .................................. 1.3-46 Table 1.3-18 Allowed GTPD Bumup and Cool Time Combinations .................................. 1.3-46 Table 1.3-19 BWR Fuel Assembly Characteristics .............................................................. 1.3-47 Table 1.3-20 BWR Fuel Assembly Loading Criteria ........................................................... 1.3-48 Table 1.3-21 BWR Fuel Assembly Loading Criteria-Enrichment Limits for 87-Assembly and 82-Assembly Configurations with Axial Blanket ................... 1.3-49 Table 1.3-22 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel Without Axial Blanket) ..................................................................... 1.3-50
- NAC International 1-iil
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Port and Coverplate The MAGNATRAN has a lid port that is closed by a bolted Type 304/304L stainless steel coverplate with dual O-rings. The four coverplate bolts are SA-193, Grade B6, Type 410 stainless steel, socket head cap screws. The bolts are countersunk flush with the top of the coverplate. The basic configuration of the lid port and coverplate includes a 5.32-inch-diameter opening to recess the coverplate and for access to the port opening and the quick-disconnect installed there. Two concentric O-rings are located on the bottom face of the coverplate, an inner metal O-ring and an outer EPDM O-ring. The inner O-ring provides the containment boundary seal for the lid port. The outer O-ring and a test port located between the two O-rings provide the means to leak test the containment boundary seal. After the leak test is completed, the seal test port is closed by a threaded plug fitted with a metal boss seal.
Lifting Trunnions and Rotation Trunnions The two lifting trunnions on the MAGNATRAN are Type 17-4 PH stainless steel, which are bolted into recesses in the top forging at diametrically opposite locations around the cask circumference. Each lifting trunnion is bolted to the top forging by nine SB637, GR N07718, nickel alloy bolts. The basic diameter of the lifting trunnions is 6.6 inches and the load-bearing width is 3.75 inches. A retainer, or flange, on the outer end of each lifting trunnion acts as a safety stop to ensure that proper engagement with the lift yoke is maintained. The MAGNATRAN lifting trunnions are designed and load tested in accordance with the requirements of ANSI N14.6 and 10 CFR 71.45(a).
Two rotation trunnions, located 17 .65 inches above the bottom of the cask and circumferentially in line with the two lifting trunnions, are offset approximately 5 inches from the cask centerline to ensure that the cask rotates in the proper direction. The rotation trunnions also serve as the cask tiedown restraint in the aft longitudinal direction. Each rotation trunnion support is Type XM-19 stainless steel housing a 17-4 PH pin and is welded to the outer shell and bottom outer forging. The neutron shield assemblies are shaped to accommodate the location and operation of the rotation trunnions.
Transport Impact Limiters The MAGNATRAN transport cask is equipped with removable impact limiters that are bolted over each end of the cask to ensure that the design impact loads for the cask are not exceeded for any of the normal conditions of transport and hypothetical accident conditions defined in 10 CFR 71.
NAC International 1.3-7
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C The lower impact limiter is bolted to the cask bottom plate by 16 equally spaced retaining rods and nuts. The upper impact limiter is similarly bolted to the cask lid.
Transportable Storage Canisters The transportable storage canister (TSC) and the closure lid are dual-certified Type 304/304L stainless steel. The canister holds the fuel basket or GTCC waste basket liner assembly and contains the contents. A schematic pf a typical TSC with a fuel basket is shown in Figure 1.3-1.
There are two different length TSCs (short and long) to accommodate the various PWR and BWR fuel assembly lengths, damaged fuel (only short), and GTCC waste (only short). The TSC body (shell and bottom) and the closure lid provide confinement, shielding and lifting capability for the TSC. The loaded TSCs include a solid stainless steel closure lid or stainless steel/carbon steel closure lid assembly with a closure ring and dual port covers to provide a dual-welded closure system. The closure lid is positioned inside the TSC on the lifting lugs above the basket assembly following fuel loading, or on the top of the GTCC waste basket liner following GTCC waste loading. After the closure lid is placed on the TSC, the TSC is moved to a workstation and the closure lid is welded to the TSC. The vent and drain ports are penetrations through the lid, which provide access for auxiliary systems to drain, dry and helium backfill the TSC. Following completion of backfilling, the dual port covers are installed and welded in each port. Removable lifting fixtures installed in the closure lid are used to lift and lower the loaded TSC. The design characteristics of the TSCs are summarized in Table 1.3-2.
The fuel TSC is designed, fabricated, tested and inspected to the requirements of the ASME Boiler and Pressure Vessel Code (ASME Code),Section III, Division 1, Subsection NB, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1.
The GTCC waste TSCs are fabricated using ASTM materials and are fabricated in accordance with ASME Code,Section III, Division 1, Subsection NF.
Criticality evaluations were performed for conditions both crediting and not crediting the TSC sealed boundary for moderator exclusion from the fissile material region. In the context of moderator exclusion, the TSC is credited with serving the 10 CFR 71.55( c) function of being a special design feature that prevents a single packing error from permitting leakage into the fissile material region. Leakage testing of the cask containment seals assures that the containment does NAC International 1.3-8
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C not leak. Regardless of credit applied to the TSC confinement boundary to prevent water in-leakage, the containment function is retained by the transport cask body.
Fuel Baskets Each TSC containing spent fuel includes a PWR, PWR-DF or BWR fuel basket that positions and supports the contents (fuel). Consistent with the TSC design, there are two different length fuel baskets (the two lengths are the same for both the PWR and the BWR fuel baskets). As described in the following sections, the design of the basket is similar for the PWR and BWR configurations. The fuel basket for each fuel type is designed, fabricated and inspected to the requirements of the ASME Code,Section III, Division 1, Subsection NG, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1.
The structural components of the PWR, PWR-DF and BWR baskets are fabricated from ASME SA537, Class 1, carbon steel. To minimize corrosion and preclude significant generation of combustible gases during fuel loading, the assembled basket is coated with electroless nickel plating using an immersion process. Following plating of the structural components, the neutron absorber panels and the stainless steel retainers are installed on the basket structure as shown on the License Drawings. The principal dimensions and materials of fabrication of the fuel basket
- and PWR damaged fuel cans are provided in Table 1.3-3 and Table 1.3-4, respectively.
The fuel basket designs minimize horizontal surfaces that could entrain water. Open paths for water flow to the drain tube and sump in the bottom of the TSC are provided. The fuel baskets are supported from the TSC bottom plate by 3-in high standoffs at the corner of the fuel tubes enabling the TSC to fill and drain evenly.
Fuel spacers may be used in the TSCs to reduce axial gaps for the spent fuel assemblies, non-fuel PWR Fuel Baskets The PWR fuel basket design is an arrangement of square fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube. The neutron absorber sheets, in conjunction with minimum TSC cavity water boron levels, provide criticality control in the basket. Each neutron absorber sheet is covered by a thin stainless steel sheet to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube. The neutron absorber NAC International 1.3-9
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C sheets may be replaced by commercial aluminum sheets on the two outside surfaces of the eight outermost fuel tubes of the PWR fuel basket (refer to sheet 3 of Drawing 71160-575). The design parameters for the two lengths of PWR fuel baskets are provided in Table 1.3-3.
Each PWR fuel basket has a capacity of up to 37 fuel assemblies in an aligned configuration.
Square tubes are assembled in an array where the tubes function as independent fuel positions and as sidewalls for the adjacent fuel positions in what is called a developed cell array.
Consequently, the 37 fuel positions are developed using only 21 tubes. The array is surrounded by side and comer weldments that serve both as sidewalls for some perimeter fuel positions and as the structural load path to the TSC shell. Each PWR basket fuel tube has a nominal 8.86-inch square opening. Each developed cell fuel position has a nominal 8.76-inch square opening.
The system is also designed to store up to four damaged fuel cans (DFCs) in the DF Basket Assembly in the short TSC. The DF Basket Assembly has a capacity ofup to 37 undamaged PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations. The arrangement of tubes and fuel positions is the same as in the standard fuel basket, but the design of each of the four comer support weldments is modified with additional structural support to provide an enlarged position for a damaged fuel can at the outermost comers of the fuel basket Each DFC location has a nominal 9.80-in square opening. A DFC or an undamaged fuel assembly may be loaded in a DFC location.
BWR Fuel Basket The BWR fuel basket design is an arrangement of square fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube for criticality control in the basket during fuel loading/unloading. Each neutron absorber sheet is covered by a sheet of stainless steel to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube.
The neutron absorber sheets may be replaced by commercial aluminum sheets on the three outer surfaces of the outermost fuel tubes of the BWR fuel basket (refer to sheet 3, Drawing 71160-599).
Each BWR fuel basket has a capacity of 87 fuel assemblies in an aligned configuration. Square tubes are assembled in an array where the tubes function as independent fuel positions and as sidewalls for the adjacent fuel positions in what is called a developed cell array. Consequently, the 87 fuel positions are developed using only 45 tubes. The array is surrounded by weldments that serve both as sidewalls for some perimeter fuel positions and as the structural load path to NAC International 1.3-10
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C the TSC shell wall. Each BWR basket fuel tube has a nominal 5.86-in square opening. Each developed cell fuel position has a nominal 5.77-in square opening.
GTCC Waste Basket Liner An ASTM A240, Type 304, stainless steel GTCC waste basket liner is designed to hold GTCC waste and dimensionally fit in a TSC. The waste basket liner design includes a shell for structural and gamma shield functions, a welded bottom plate, and lifting lugs welded on the inside diameter of the shell so that the liner may be loaded with GTCC waste prior to being inserted into a TSC (Table 1.3-5). The liner design also includes an outer ring and a middle support under the bottom plate and drain holes in the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC includes a sump location in the bottom plate and the closure lid includes a drain tube assembly to enable draining and drying of the loaded TSC.
The GTCC waste basket liner and TSC are designed, fabricated and inspected in accordance with ASME Code,Section III, Division 1, Subsection NF. The lifting features of the GTCC components are designed for noncritical lifting in accordance with NUREG-0612 and ANSI N14.6, with safety factors of 3 on material yield strength and 5 on material ultimate strength
- applied .
Damaged Fuel Can The MAGNASTOR Damaged Fuel Can (DFC), shown in Figure 1.3-3, is provided to accommodate damaged PWR fuel assemblies. The DFC may also contain PWR fuel assemblies in an undamaged condition or fuel debris equivalent to, or less than, one PWR fuel assembly.
The primary function of the DFC is to confine the fuel material within the can to minimize the potential for dispersal of the fuel material into the TSC cavity. In normal operation, the DFC is in a vertical orientation.
The DFC is fabricated from Type 304 stainless steel and has an 8.7-in square inside dimension (see Figure 1.3-3). The DFC is designed in two lengths: an overall length of 166.9 inches with a nominal cavity length of 164.0 inches; or an overall length of 171.8 inches with a nominal cavity length of 169.0 inches (shorter fuel assemblies may be accommodated with a fuel assembly spacer to limit axial movement). For the shorter DFC, a DFC spacer is used in the DF basket assembly or alternatively fixed to the DFC bottom plate to provide an overall height ofDFC and spacer of 171.5 inches. The side plates that form the upper end of the DFC are 0.15-in thick and the tube body walls are 0.048-in thick (18-gage sheet). The DFC lid plate and bottom
- NAC International 1.3-11
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C thicknesses total 11/16 (0.688) inch and the lid overall height is 2.32 inches. The DFC bottom plate thickness is 5/8 (0.625) inch. The DFC lid and bottom include screened drain holes. The DFC is designed, fabricated, tested and inspected to the requirements of the AS11E Code,Section III, Division 1, Subsection NG, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1.
Cask Cavity Spacer An ASME SA240, Type 304, stainless steel cask cavity spacer is used in the upper end of the MAGNATRAN cavity to limit the axial movement of the short TS Cs. The spacer consists of six concentric rings welded to a flat plate. The depth of the rings, i.e., the length of the spacer, is approximately seven inches, which represents the difference in length between the short and long TS Cs. The spacer is bolted through the flat plate to the underside of the cask lid.
Containment System MAGNATRAN provides a containment system to retain the radioactive material and gas contents during transport operations. The cask design pressure is 120 psig. The MAGNATRAN containment system components include the bottom inner forging, inner shell, top forging, cask lid and lid bolts, metal inner O-ring Oid), coverplate and bolts, and metal inner O-ring (coverplate). The cask lid is sealed by two concentric O-rings, as is the coverplate for the lid port. In both cases, the metal inner O-ring is the containment boundary and the outer EPDM O-ring forms the annulus to facilitate leakage testing of the containment seal following installation of the lid and coverplate. A test port is provided in each annulus to enable the performance of the leakage tests. After the leakage tests are completed, each test port is closed by a stainless steel plug and boss seal. A sketch of the containment boundary is shown in Figure 1.3-2. All of the containment boundary components are defined on the License Drawings in Section 1.4-3.
NAC International 1.3-12
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 1.3.2 Contents The MAGNASTOR fuel TSC or Greater Than Class C (GTCC) TSC can be transported in the MAGNATRAN.
The MAGNASTOR fuel TSC design capacity is up to 37 undamaged uranium PWR fuel assemblies in the 37 PWR basket assembly, or up to 87 undamaged uranium BWR fuel assemblies in the 87 BWR basket assembly. The fuel TSC with the DF basket assembly has a capacity ofup to 37 undamaged uranium PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations as shown in Figure 1.3-4.
Fuel assemblies are assigned to two groups of PWR and two groups ofBWR fuel assemblies on the basis of fuel assembly length. Fuel assemblies are restricted to those with zirconium alloy-clad fuel rods; no steel-clad assemblies are considered. PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC. BWR fuel assemblies containing channels may be loaded in the TSC.
The initial enrichment limits are shown in Table 1.3-6 and Table 1.3-19 for PWR and BWR fuel, respectively, and exclude the loading of fuel assemblies enriched to less than 1.3 wt% 235 U, including unenriched fuel assemblies. Upper enrichment limits in these tables are based on moderator intrusion into the TSC. Crediting the TSC with moderator exclusion allows 5 wt.%
235 U enriched fuel to be loaded up to maximum capacity regardless of fuel type or basket absorber content. The evaluated dry system reactivity is independent of fuel physical parameters with the exception of maximum fissile mass. Fuel assemblies containing low enriched, unenriched, and/or annular axial end-blankets may be loaded into the TSC. The end blankets are typically regions oflow enriched or natural uranium oxide. Unenriched (natural uranium oxide) end blankets are limited to a nominal length no greater than 6 inches (dimension applicable to top and bottom of the fuel rod, not cumulative).
BWR/2,3 fuel assemblies with a fuel assembly length of approximately 171 inches will be loaded into the short TSC (nominal length of 184.75 inches). BWR/4-6 fuel assemblies with a fuel assembly length of approximately 176 inches will be loaded into the long TSC (nominal length 191.75 inches). PWR fuel assemblies, with the exception of fuel identified as type CE 16x 16, in Table 1.3-6 must be loaded into a short TSC (i.e., be loaded into a transport configuration with a canister spacer). CE16 fuel may be loaded into a either short or long TSC.
An exception to the PWR CE16 guideline is the System 80 16x16 fuel with an active fuel length of 150 inches whose overall length exceeds the short TSC cavity length and, therefore, must be loaded into the long TSC.
Empty fuel rod positions are filled with a solid filler rod or a solid neutron absorber rod that
- displaces a volume not less than that of the original fuel rod.
NAC International 1.3-21
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C A MAGNASTOR TSC waste consisting of solid, irradiated, contaminated hardware, nonfuel-bearing carbon and/or stainless steel material (e.g., loose nuts and bolts, etc.) provided that the bounding weight of the loaded TSC as shown in Table 2.1.3-1 is not exceeded.
Greater Than Class C (GTCC) waste is defined in 10 CFR 61 .55(a)(3) and (4) by the concentration of long-lived radionuclides, i.e., 14C, 59Ni, and 94NB, and/or short-lived radionuclides, i.e., 3H, 6°Co, and 63Ni. The disposal of GTCC waste is controlled by 10 CFR 61.
GTCC waste consists of radiation activated and surface contaminated steel, and/or plasma cutting debris (dross). Stainless steel core baffle structure - baffle plates and angles, baffle formers, and lower core plates, located adjacent to the reactor vessel in a high neutron flux field, is the major component of GTCC waste.
The GTCC waste to be transported in the MAGNATRAN is placed in a GTCC waste basket liner, which is loaded into a GTCC TSC. The GTCC TSC and welded closure lid are geometrically identical to that of the fuel TSC.
PWR Fuel The PWR fuel evaluations are based on bounding PWR fuel assembly parameters that maximize the source terms for the shielding evaluations, the reactivity for criticality evaluations, the decay heat load for the thermal evaluations, and the fuel weight for the structural evaluations. These bounding parameters are selected from the various spent fuel assemblies that are candidates for loading in the TSC. The bounding fuel assembly values are established based primarily on how the principal parameters are combined, and on the loading conditions (or restrictions) established for a group of fuel assemblies based on its parameters.
The limiting parameters of the PWR fuel assemblies authorized for loading in the TSC are shown in Table 1.3-6. The maximum initial enrichment authorized represents the peak fuel rod enrichment for variably enriched PWR fuel assemblies. The PWR fuel assembly characteristics are summarized by fuel assembly type in Chapter 6, with burnup credit loading curves correlating maximum initial enrichment for a given assembly average burnup at various absorber sheet loading, also listed in Chapter 6. Table 1.3-6 assembly physical information is limited to the analysis input of fuel mass, array configuration, and number of fuel rods. These analysis values are key inputs to the shielding and criticality (water moderated) evaluations in Chapters 5 and 6, respectively. Lattice parameters dictating system reactivity are detailed in Chapter 6.
Enrichment limits are set for each fuel type to produce reactivities at the upper subcritical limit (USL).
NAC International 1.3-22
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C When crediting moderator exclusion lattice parameters are not significant to demonstrating subcriticality of the package. Critical parameters are limited to enrichment and fuel mass.
The maximum TSC decay heat load for the transport of PWR fuel assemblies is 23.0 kW. The uniform loading pattern permitted limits assemblies to a maximum heat load of 0.622 kW/assembly. Neutron absorber with Type 2 thermal conductivity (see Table 3.2-12) is required for PWR basket with a maximum heat load of 23kW. For the PWR basket with neutron absorbers with Type 1 thermal conductivity (see Table 3.2-12), the heat load is limited to 22 kW, with the individual assembly decay heat limited to 0.595 kW. The bounding thermal evaluations are based on the Westinghouse 17xl 7 fuel assembly. The minimum cool times are determined based on the maximum decay heat load of the contents and meeting transport dose limits. The fuel assemblies and source terms that produce the maximum dose rates are summarized in Chapter 5.
The DF basket assembly configuration for PWR fuel with damaged fuel can locations is shown in Figure 1.3-4. Each DFC may contain an undamaged PWR fuel assembly, a damaged PWR fuel assembly, or PWR fuel debris equivalent to one PWR fuel assembly. Undamaged PWR fuel assemblies may be placed directly in the DFC locations of a DF Basket Assembly. PWR fuel assemblies loaded in a DFC shall not contain nonfuel hardware, with the exception of instrument tube tie components and steel inserts.
A PWR fuel assembly weight of 1,680 pounds based on a B&W 15x15 fuel assembly with control components inserted, has been structurally evaluated in each location of the PWR fuel basket, equaling a total contents weight of 62,160 lbs. A bounding weight of 1,814 pounds is evaluated for each loaded damaged fuel can in the damaged fuel configuration of the PWR DF fuel basket. A total contents weight of 61,184 lbs is specified for the PWR DF basket to limit the maximum loaded TSC weight to 104,500 lbs. The analyzed contents weight provides the most significant measure of the basket performance. Accordingly, a 5% increase in the maximum weight per undamaged fuel location of 1,765 lbs is permitted while maintaining a maximum contents weight consistent with the basket evaluation.
As noted in Table 1.3-6, PWR fuel assemblies may include nonfuel hardware placed into the fuel assembly guide tubes and/or instrument tube. Nonfuel hardware that is located in the active fuel region is referred to as inserts in this SAR. Nonfuel components, such as thimble plugs, may not reach into the active fuel region and do not have a significant effect on system reactivity.
Westinghouse 15 x 15 PWR fuel assemblies that have been enhanced to address top nozzle stress corrosion cracking may use nonfuel hardware to prevent the separation of the top nozzle .
- NAC International 1.3-23
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C BWR Fuel The BWR fuel evaluations are based on bounding BWR fuel assembly parameters that maximize the source terms for the shielding evaluations, the reactivity for the criticality evaluations, the decay heat load for the thermal evaluations, and the fuel weight for the structural evaluations.
These bounding parameters are selected from the various spent fuel assemblies that are candidates for loading in the TSC. The bounding fuel assembly values are established based primarily on how the principal parameters are combined, and on the loading conditions or restrictions established for a group of fuel assemblies based on its parameters. Each TSC may contain up to 87 undamaged BWR fuel assemblies. To increase allowed assembly enrichments over those determined for the 87-assembly basket configuration, an optional 82-assembly loading pattern may be used. The required fuel assembly locations in the 82-assembly pattern are shown in Figure 1.3-5.
The limiting parameters of the BWR fuel assemblies authorized for loading in the TSC are shown in Table 1.3-19. The maximum initial enrichment represents the peak planar-average enrichment. The BWR fuel assembly characteristics are summarized by fuel type in Chapter 6.
Table 1.3-19 assembly physical information is limited to the critical analysis input of fuel mass, array configuration, and number of fuel rods. These analysis values are key inputs to the shielding and criticality (water moderated) evaluations in Chapters 5 and 6. Lattice parameters dictating system reactivity are detailed in Chapter 6. Enrichment limits are set for each fuel type to produce reactivities at the USL. The maximum decay heat load per TSC for the transport of BWR fuel assemblies is 22.0 kW (0.253 kW/assembly). Only uniform loading is permitted for BWR fuel assemblies. The bounding thermal evaluations are based on the GE 1Ox10 fuel assembly. The minimum cooling times are determined based on the maximum decay heat load of the contents and meeting transport dose limits.
BWR fuel assemblies may contain partial-length fuel rods. Table 1.3-20 contains the type of BWR assemblies and the number of partial-length rods included in the analysis in this SAR.
Locations for the partial-length rods within the lattice are illustrated in Figure 1.3-8.
When crediting moderator exclusion lattice parameters, including presence or absence of partial length rods, are not significant to demonstrating subcriticality of the package. Critical parameters are limited to enrichment and fuel mass.
A bounding BWR fuel assembly weight of704 pounds based on the maximum weight of GE 7x7 and 8x8 assemblies with channels has been structurally evaluated in each storage location of the BWR basket as well as the two additional locations coinciding with the drain and vent ports, equaling a total contents weight of 62,656 lbs. The analyzed contents weight provides the most significant measure of the basket performance. Accordingly, a 5% increase in the maximum NAC International 1.3-24
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C weight per fuel location of 739 lbs is permitted while maintaining a maximum contents weight consistent with the basket evaluation.
As noted in Table 1.3-19, the evaluation ofBWR fuel envelops unchanneled assemblies and assemblies with channels up to 120 mils thick.
GTCC Waste The GTCC waste to be transported in the MAGNA TRAN transport cask consists of sections of core baffie plates and angles, baffie formers, lower core plates and miscellaneous related hardware associated with these components. The major components are cut into pieces of a size that are loaded into a GTCC waste basket liner. Small residual pieces of GTCC may be loaded into stainless steel strainer baskets for handling. The loaded strainer baskets are stacked in a stainless steel pipe cell that has been placed in the GTCC waste basket liner to retain spacing prior to GTCC initial loading. Any dross material (fines and debris) generated by the cutting operations will be disposed of as low-level radioactive or GTCC waste.
Each GTCC waste basket liner may contain up to 55,000 pounds of GTCC waste including the weight of strainer baskets and pipe spacers. The GTCC waste basket liner is transported in a GTCC TSC with a welded closure lid. The GTCC waste basket liners have twelve 1.0-inch
- diameter holes in the bottom plate, and outer ring and middle supports under the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC has a sump in the bottom plate, and the closure lid includes a drain tube assembly to enable draining and vacuum drying of the loaded TSC. Consequently, no hydrogen generation occurs as a result of residual water.
The radionuclide composition of the waste was determined based on radiochemical assay of samples and dose rate measurements. The isotope that primarily contributes to the radiological 60 source term is Co. The source terms applied in the evaluation of the GTCC waste are presented in Chapter 5 of this SAR.
Fuel and GTCC Content Limits Spent fuel and GTCC waste shipments in the MAGNA TRAN shall be subject to the following limits:
- 1. The maximum contents weight for the MAGNA TRAN transport cask shall not exceed 106,000 pounds.
- 2. The design basis fuel characteristics shall be in accordance with Table 1.3-6 and Table 1.3-19.
- 3. The total decay heat of the cask cavity contents shall not exceed:
- a. 23 kW for PWR fuel with a uniform loading pattern
- NAC International 1.3-25
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- b. 22 kW for PWR fuel loaded in a basket with neutron absorbers having Type 1 thermal conductivity (see Table 3.2-12)
- c. 22 kW for BWR fuel
- d. GTCC waste content, the decay heat limit is 1.7 kW.
- 4. The total weight of the PWR fuel assemblies in the TSC, including standard nonfuel hardware and spacers (if used), shall not exceed 62,160 pounds.
- 5. The total weight of the PWR fuel assemblies in the DF PWR TSC, including standard nonfuel hardware and spacers (if used), shall not exceed 61,184 pounds.
shall not exceed 62,656 pounds.
- 7. GTCC waste consists of solid, irradiated, and contaminated hardware provided the quantity of fissile material does not exceed a Type A quantity and does not exceed the mass limits of 10 CFR 71.15.
The specific Curie content source of the GTCC shall be limited to:
- a. a maximum of2.7 Ci 60 Co/lb averaged over GTCC contents
- b. a localized peak 16.1 Ci 60 Co/lb
- c. a total 60 Co activity of 85,760 Ci at transport.
The maximum allowed weight of this waste is 55,000 lbs.
- 8. Any number ofMAGNATRAN casks may be shipped at one time by rail, ship, barge or heavy-haul vehicle with the exception of a PWR-DF basket with DFC which requires only one cask to be shipped at one time.
- 9. Radiation levels shall not exceed the requirements of 10 CFR 71.47 and 10 CFR 71.51 for a closed transport vehicle.
- 10. Surface contamination levels shall not exceed the requirements of 10 CFR 71.87(i)(l).
- 11. Cask contents transported in a TSC with a PWR fuel basket shall be uranium undamaged PWR fuel assemblies in accordance with the limiting values shown in Table 1.3-6 and Table 1.3-7 and shall meet the following specifications:
- a. Zirconium-based alloy cladding.
- b. Enrichment, post-irradiation cooling time and burnup credit load curves in accordance with Tables 1.3-6, 1.3-8 through 1.3-11, and Figure 1.3-6 (burnup credit curves are only applicable to systems not crediting moderator exclusion) with moderator exclusion up to 5wt% fuel may be loaded for all fuel types.
- c. Maximum assembly average burnup shall be :S 60,000 MWd/MTU.
- d. Decay heat per fuel assembly: 622 watts (includes non-fuel hardware contribution). For the PWR basket with neutron absorbers with Type 1 thermal conductivity (see Table 3.2-12), the decay heat per fuel assembly is limited to 595 watts.
NAC International 1.3-26
MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- e. Nominal fresh fuel dimensions: assembly length (in.) :S 178.3 assembly width (in.) :S 8.54
- f. Fuel assembly weight (lbs.): :S 1,765 (including nonfuel hardware and fuel spacers)
- g. Spent fuel contents shall be loaded in accordance with the loading tables in Chapter 5, Section 5.8.3, of this SAR.
- h. Quantity per TSC: up to 37 undamaged PWR fuel assemblies shown in Figure 1.3-6.
Figure 1.3-6 indicates the fuel storage locations that shall be empty, at a minimum, when implementing the 36, 35 and 33 loading patterns for burnup credit purposes.
- 1. Undamaged PWR fuel assemblies may contain nonfuel hardware (NFHW). Fuel assembly lattices not containing the nominal number of fuel rods specified in Table 1.3-7 must contain solid filler rods that displace a volume equal to, or greater than, that of the fuel rod that the filler rod replaces. Fuel assemblies may have stainless steel rods inserted to displace guide tube "dashpot water. Nonfuel hardware cool times shall be in accordance with Tables 1.3-16 through 1.3-18. Alternatively, the 6°Co curie limits in Table 1.3-17 and Table 1.3-18 may be used to establish site-specific nonfuel hardware constraints. Note that fuel assemblies defined as CE14 and CE16 are not allowed to contain BPRA or TP type nonfuel hardware .
J. Fuel spacers may be used in the TSCs to reduce axial gaps for the spent fuel assemblies and non-fuel hardware.
- k. Unenriched and unirradiated (i.e., not inserted in-core) fuel assemblies are not authorized for loading. Unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than six inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition.
I. Reactor control components (RCC) are restricted to fuel storage locations No. 11, 12, 13, 18, 19, 20, 25, 26 and 27 (Figure 1.3-6). Minimum RCC cool times are:
Minimum Cool Time Maximum Exposure (years) (GWd/MTU) 10 180 14 270 20 360 Interpolation is not allowed between data points.
- m. One Neutron Source or Neutron Source Assembly (NSA) is permitted to be loaded in a TSC in fuel storage locations No. 11, 12, 13, 18, 19, 20, 25, 26 or 27 (Figure 1.3-6).
- Neutron source assemblies may contain source rods attached to hardware similar in NAC International 1.3-27
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. For NSAs containing absorber rodlets, the BPRA cool time and burnup/exposure or hardware 60 Co curie limit listed in Table 1.3-17 are applied to the neutron sources. NSAs having only thimble plug rodlets require the thimble plug restriction in Table 1.3-18 to be applied.
Combination NSAs, containing both thimble plug and burnable absorber rodlets must apply the more limiting of the two minimum cool time/curie limit. Fuel assemblies loaded with the NSAs must apply the additional cool times listed in Table 1.3-16. Fuel types indicated as CE14 and CE16 are not permitted to be loaded with NSAs.
- n. Fuel assemblies may contain any number of unirradiated (i.e., not inserted in-core) nonfuel solid filler fuel replacement rods. Activated stainless steel rods are limited to five per assembly, one assembly per basket, at a maximum steel rod burnup/exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless steel rods must be cooled either a minimum of 21 years or the Section 5.3 loading table minimum cool time plus one year, whichever is greater.
- o. Westinghouse fuel assemblies may contain a hafnium absorber assembly (HFRA) at a maximum burnup/exposure of 4.0 GWd/MTU and a minimum cool time of 16 years .
Fuel assemblies loaded with an HFRA must apply the additional cool times listed in Table 1.3-16.
- p. Under-burned (assemblies with burnup less than that dictated by the burnup credit loading curve) Westinghouse 15 x 15 PWR fuel assemblies may be loaded provided that they include Ag-In-Cd full-length RCCAs and are loaded in the basket locations that RCCs are allowed (see item I for RCCA loading). Burnup must be greater than or equal to 12,000 MWd/MTU. Enrichment must be equal to or less than 4.05 wt.% 235 U. The basket must include absorber sheets with an effective 10 B areal density of 0.036 g/cm 2 .
For the loading of low burnup fuel, the RCCAs must be full length (i.e. spider component included). RCCA exposure must be equal to or less than 200,000 MWd/MTU. Any assemblies loaded without an RCCA inserted must meet the burnup credit loading curve for the applicable assembly loading profile.
Burnup credit curves and the need to insert RCCAs for criticality control are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt. %
235 U, with no burnup requirement, is permitted when crediting moderator exclusion.
- 12. Cask contents transported in a TSC with a DF Basket Assembly shall be uranium undamaged PWR fuel assemblies and damaged fuel (damaged PWR fuel assemblies or PWR fuel debris)
NAC International 1.3-28
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C in accordance with the limiting values shown in Table 1.3-6 and Table 1.3-7 and shall meet the following specifications:
- a. Zirconium-based alloy cladding.
- b. For the 33 non-DFC fuel locations in the DF Basket Assembly, enrichment, post-irradiation cooling time and bumup credit load curves in accordance with Tables 1.3-6, 1.3-12 through 1.3-15, and Figure 1.3-6 for a TSC with a DF Basket Assembly containing DFCs. For a TSC with a DF Basket Assembly that does not contain any DFCs, the enrichment, post-irradiation cooling time and bumup credit load curves in accordance with Tables 1.3-6, 1.3-8 through 1.3-11, and Figure 1.3-4 may be used for all fuel locations (bumup credit curves are only applicable to systems not crediting moderator exclusion).
- c. For the up to four DFC locations in a DF Basket Assembly containing damaged fuel, the damaged fuel shall have a minimum bumup of 5 GWd/MTU, a maximum enrichment of 4.05 wt% 235 U, and a minimum cool time of 15 years. Application of moderator exclusion allows increasing the maximum initial enrichment to 5 wt. % 235 U, with no bumup requirement.
- d. Maximum assembly average bumup shall be :S 60,000 MWd/MTU .
- e. Decay heat per fuel assembly: 622 watts (590.5 watts for bumup > 45,000 MWd/MTU, includes non-fuel hardware contribution). For the PWR basket with neutron absorbers with Type 1 thermal conductivity (see Table 3.2-12), the decay heat per fuel assembly is limited to 595 watts
- f. Nominal fresh fuel assembly: length (in.) :S 167.0
- g. Nominal fresh fuel assembly: width (in.) :S 8.54
- h. Fuel assembly weight (lbs.): :S 1,765 (including nonfuel hardware, DFCs and fuel spacers)
- 1. Spent fuel contents shall be loaded in accordance with the loading tables in Section 5.8.3 with additional cool time for damaged fuel found in Table 5.8-49 of this SAR. The additional cool time from Table 5.8-49 applies to all assemblies loaded in a damaged fuel TSC with damaged fuel.
J. Quantity per TSC: Up to a total of 37 undamaged PWR fuel assemblies, including up to four DFCs containing undamaged PWR fuel assemblies, damaged PWR fuel assemblies, and/or PWR fuel debris loaded in DFC location Nos. 4, 8, 30 and 34, as shown on Figure 1.3-4, for the DF Basket Assembly. Figure 1.3-6 indicates the fuel storage locations that
- NAC International 1.3-29
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C shall be empty, at a minimum, when implementing the 36, 35 and 33 loading patterns for bumup credit purposes.
- k. The contents of a DFC must be less than, or equivalent to, one undamaged PWR fuel assembly. PWR fuel assemblies loaded in a DFC shall not contain nonfuel hardware with the exception of instrument tube tie components, guide tube anchors or other similar devices, and steel inserts.
I. Undamaged PWR fuel assemblies not loaded in a DFC may contain nonfuel hardware consistent with Table 1.3-16. Fuel assembly lattices not containing the nominal number of fuel rods specified in Table 1.3-7 must contain solid filler rods that displace a volume equal to, or greater than, that of the fuel rod that the filler rod replaces. Fuel assemblies may have stainless steel rods inserted to displace guide tube "dashpot" water. Nonfuel hardware cool times shall be in accordance with Tables 1.3-16 through 1.3-18.
Alternatively, the 60 Co curie limits in Tables 1.3-17 and 1.3-18 may be used to establish site-specific nonfuel hardware constraints. Note that fuel assemblies defined as CE14 and CEl 6 are not allowed to contain BPRA or TP type nonfuel hardware.
- m. Fuel spacers may be used in the TSCs to reduce axial gaps for the spent fuel assemblies, non-fuel hardware or damaged fuel cans.
- n. Unenriched and unirradiated (i.e., not inserted in-core) fuel assemblies are not authorized
- for loading. Unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than six inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition.
- o. Reactor control components (RCC) are restricted to fuel storage location Nos. 11, 12, 13, 18, 19, 20, 25, 26 and 27 (Figure 1.3-4). Minimum RCC cool times are:
Minimum Cool Time Maximum Exposure (years) (GWd/MTU) 10 180 14 270 20 360 Interpolation is not allowed between data points.
- p. One Neutron Source or Neutron Source Assembly (NSA) is permitted to be loaded in a TSC in fuel storage location Nos. 11, 12, 13, 18, 19, 20, 25, 26 or 27 (Figure 1.3-4).
Neutron source assemblies may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. For NSAs NAC International 1.3-30
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C containing absorber rodlets, the BPRA cool time and burnup/exposure or hardware 6°Co curie limit listed in Table 1.3-17 are applied to the neutron sources. NSAs having only thimble plug rodlets require the thimble plug restriction in Table 1.3-18 to be applied.
Combination NSAs, containing both thimble plug and burnable absorber rodlets must apply the more limiting of the two minimum cool time/curie limit. Fuel assemblies loaded with the NSAs must apply the additional cool times listed in Table 1.3-16. Fuel types indicated as CE14 and CE16 are not permitted to be loaded with NSAs.
- q. Fuel assemblies may contain any number ofunirradiated (i.e., not inserted in-core) nonfuel solid filler fuel replacement rods. Activated stainless steel rods are limited to five per assembly, one assembly per basket, at a maximum steel rod burnup/exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless steel rods must be cooled either a minimum of21 years or the item 12.i indicated minimum cool time plus one year, whichever is greater.
- r. Westinghouse fuel assemblies may contain a hafnium absorber assembly (HFRA) at a maximum burnup/exposure of 4.0 GWd/MTU and a minimum cool time of 16 years.
Fuel assemblies loaded with an HFRA must apply the additional cool times listed in Table 1.3-16.
- s. Under-burned (assemblies with burnup less than that dictated by the bumup credit loading curve) Westinghouse 15x15 PWR fuel assemblies may be loaded provided that they include Ag-In-Cd full-length RCCAs and are loaded in the basket locations that RCCs are allowed (see item o for RCCA loading). Burnup must be greater than or equal to 12,000 MWd/MTU. Enrichment must be equal to or less than 4.05 wt.% 235 U. The basket must include absorber sheets with an effective 10B areal density of 0.036 g/cm2
- For the loading oflow burnup fuel, the RCCAs must be full length (i.e. spider component included). RCCA exposure must be equal to or less than 200,000 MWd/MTU. Any assemblies loaded without an RCCA inserted must meet the burnup credit loading curve for the applicable assembly loading profile. Burnup credit curves and the need to insert RCCAs for criticality control are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt.% 235 U, with no burnup requirement, is permitted when crediting moderator exclusion.
- t. Damaged CE 16x16 fuel assemblies are not to be loaded in the MAGNATRAN system.
- 13. Cask contents transported in a TSC with a BWR fuel basket shall be uranium undamaged BWR fuel assemblies in accordance with the limiting values shown in Table 1.3-19 and Table 1.3-20 and shall meet the following specifications:
- a. Zirconium-based alloy cladding .
NAC International 1.3-31
MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- b. Enrichment, post-irradiation cooling time and average assembly burnup in accordance with Tables 1.3-19, 1.3-21, and 1.3.-22 and Figures 1.3-5, 1.3-7 and 1.3-8.
- c. Decay heat per fuel assembly: uniform loading 253 watts
- d. Nominal fresh fuel dimensions: assembly length (in.) :S 176.2
- e. Assembly width (in.) :S 5 .52
- f. Fuel assembly weight (lbs.) :S 739 lbs (including channel and fuel spacers) with a maximum contents weight of 62,656 lbs.
- g. Spent fuel contents shall be loaded in accordance with the loading tables in Chapter 5, Section 5.8.4, of this SAR.
- h. Quantity per TSC: up to 87 undamaged BWR fuel assemblies as shown in Figure 1.3-7.
- 1. Allowable fuel assembly locations for the 82-assembly BWR fuel basket configurations are shown in Figure 1.3-5 (location numbering for the 82-assembly basket is the same as that shown for the 87-assembly basket in Figure 1.3-7).
J. Prior to use of the 82-assembly configuration, the center cell weldment and upper weldments with blocking strap must be in place to physically block the designated nonfuel locations (shown in Figure 1.3-5). Less than 82 assemblies may be loaded when implementing the 82-Assembly configuration provided the required fuel storage locations are empty, at a minimum.
The 82-Assembly configuration is the result of criticality constraints on maximum enrichment. When crediting moderator exclusion this configuration is not required as full capacity (87-Assembly) is permitted at an initial enrichment up to 5 wt.% 235 U.
- k. BWR fuel assemblies may be unchanneled, or channeled with zirconium-based alloy channels.
I. BWR fuel assemblies with stainless steel channels are not authorized.
- m. Fuel assembly lattices not containing the assembly type-specific nominal number of fuel rods specified in Table 1.3-20 must contain solid, unirradiated, filler rods that displace a volume equal to, or greater than, that of the fuel rod that the filler rod replaces
- n. Spacers may be used in the TSCs to fill axial gaps and provide support for the spent fuel assemblies.
- o. Unenriched and unirradiated (i.e., not inserted in-core) fuel assemblies are not authorized for loading. Unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than six inches.
NAC International 1.3-32
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Figure 1.3-8 BWR Partial Length Fuel Rod Location Sketches 000000000 0000000000 QxQQxQQxO oxoooxooxo 000000000 0000000000 00000000 0000000000 000 ooxo 0 xoo ox 0 oxo 0000 000 0000 000 0000 000000000 0000000000 oxooxooxo 0x00x000x0 000000000 0000000000 0
- Fuel Rod Locallon 0 = Fuel Rod Locallon X c Parual Rod Location X = Par ral Rod Locallon B9_74A 8 Partial Length Rods B10_91A 8 Partial Length Rods
- 0000000000 oxoxooxoxo 0000000000 oxoooooxo 00000 00000 oxoxo oxoxo 00000 00000 oxooo oooxo oooox 000 0000 0000 0000 xoooo 0000 0000 oxo oooxo oxooo oooxo 0000000000 00000 00000 OxOxOOxOxO oxoxo oxoxo 0000000000 00000 00000 0 = Fuel Rod Locallon 0 = Fuel Rod Locallon X = Partial Rod Locallon X "Partral Rod Locallon BJ O_92A I 4 Partial Length Rods Bl O_96A I 2 Partial Length Rods
- NAC International 1.3-37
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 1.3-6 PWR Fuel Assembly Characteristics Characteristic Fuel Class 14x14 14x14 15x15 15x15 16x16 17x17 Base Fuel Typea CE,SPC W,SPC W,SPC BW, FCF CE BW, SPC, W, FCF Max Initial Enrichment (wt% 235 U) 5.0 5.0 5.0 5.0 5.0 5.0 2
Min Initial Enrichment (wt% 35LJ) 1.3 1.3 1.3 1.3 1.3 1.3 Number of Fuel Rods 176 179 204 208 236 264 Max Assembly Average Bumup (MWd/MTU) 60,000 60,000 60,000 60,000 60,000 60,000 Min Cool Time (years) 4 4 4 4 4 4 Max Weight (lb) per Storage Location See Note 1 See Note 1 See Note 1 See Note 1 See Note 1 See Note 1 Max Decay Heat (Watts) per Fuel Location See Note 2 See Note 2 See Note 2 See Note 2 See Note 2 See Note 2
- Fuel cladding is a zirconium-based alloy.
- All reported enrichment values are nominal preirradiation fabrication values.
- Weight includes the weight of nonfuel-bearing components.
- Assemblies may contain nonfuel hardware and/or fuel replacement rods (also referred to as filler rods). Filler rods are considered to be a component of spent nuclear fuel assemblies and not nonfuel hardware. Filler rods may be burnable absorber rods, stainless steel rods or zirconium alloy rods.
- PWR fuel may be loaded using bumup credit. Maximum enrichment is as a function of minimum bumup as specified in Chapter 6. Maximum initial enrichment represents the peak fuel rod enrichment for variably-enriched fuel assemblies.
- Spacers may be used to axially position fuel assemblies to facilitate handling.
Notes:
- 1. Maximum weight per storage location is 1,765 lbs (including nonfuel hardware, DFCs and fuel spacers) with a maximum contents weight of 62,160 lbs for the PWR basket and 61,184 lbs for the DF basket.
- 2. For PWR baskets with Type 2 thermal conductivity neutron absorbers, the maximum heat load is 622 watts per storage location and PWR baskets with Type 1 thermal conductivity neutron absorbers the maximum heat load is 595 watts per storage location.
a Indicates assembly and/or nuclear steam supply system (NSSS) vendor/type referenced for fuel input data Fuel acceptability for loading is not restricted to the indicated vendor provided that the fuel assembly meets the load limits. Abbreviations are as follows: Westinghouse (W),
Combustion Engineering (CE), Siemens Power Corporation (SPC), Babcock and Wilcox (BW), and Framatome Cogema Fuels (FCF).
NAC International 1.3-38
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Table 1.3-11 Maximum Initial Enrichment - Undamaged Fuel Configuration WE15 - Optional Configurations Zero (0) Max Initial Enrichment (wt % 235 U) 8umup = C4 x 8umup (GWd/MTU) + Cs
- Maximum. 8umup (GWd/MTU) 18 S 8urnup 8umup (GWd/MTU)
Assemblies Enrichment < 18 (GWd/MTU) S 30 > 30 (wt% 235LJ) C4 I Cs C4 I Cs C4 Cs 0.036 g/cm2 10 8 Absorber 36 2.0 0.0497 1.93 0.0681 1.99 0.0747 2.00 35 2.1 0.0507 1.97 0.0673 2.08 0.0730 2.12 33 2.2 0.0504 2.12 0.0664 2.29 0.0745 2.32 0.030 g/cm2 10 8 Absorber 36 2.0 0.0494 1.87 0.0687 1.90 0.0737 1.93 35 2.0 0.0499 1.92 0.0688 1.97 0.0740 1.99 33 2.1 0.0497 2.06 0.0686 2.15 0.0724 2.29 0.027 g/cm2 10 8 Absorber 36 2.0 0.0501 1.83 0.0677 1.87 0.0741 1.84 35 2.0 0.0494 1.89 0.0675 1.94 0.0735 1.96
- 33 2.1 0.0492 2.03 0.0674 2.12 0.0730 2.21
- NAC International 1.3-43
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 1.3-12 Maximum Initial Enrichment -PWR Damaged Fuel Configuration - 0.036 g/cm 2 1°B Absorber Max Initial Enrichment (wt % 235 U)
Zero (0) = C4 x Burnup (GWd/MTU) + Cs Assembly Burnup Burnup 18 S Burnup 30 < Bumup 50 <
ID Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Bumup (wt%) <18 S30 S50 (GWd/MTU)
C4 Cs C4 Cs C4 Cs C4 Cs BW 15x15 1.6 0.0453 1.42 0.0681 1.29 0.0750 1.03 0.0750 0.736 BW 17x17 1.6 0.0476 1.45 0.0668 1.37 0.0712 1.17 0.0712 0.891 CE 14x14 1.9 0.0504 1.79 0.0696 1.75 0.0751 1.60 0.0751 1.60 CE 16x16 1.9 0.0484 1.79 0.0679 1.74 0.0758 1.52 0.0758 1.52 WE 14x14 1.9 0.0542 1.85 0.0729 1.85 0.0794 1.75 0.0794 1.75 WE 15x15 1.6 0.0482 1.43 0.0692 1.27 0.0738 1.08 0.0738 0.767 WE 17x17 1.6 0.0439 1.45 0.0657 1.35 0.0732 1.00 0.0732 0.700 Table 1.3-13 Maximum Initial Enrichment - PWR Damaged Fuel Configuration - 0.030 g/cm 2 1°B Absorber Max Initial Enrichment (wt % 235 U)
Zero (0) = C4 x Burnup (GWd/MTU) + Cs Assembly Burnup Bumup 18 S Burnup 30 < Bumup 50 <
ID Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Bumup (wt%) < 18 :S 30 S50 (GWd/MTU)
C4 Cs C4 C5 C4 Cs C4 C5 BW 15x15 1.5 0.0487 1.31 0.0660 1.26 0.0740 0.896 0.0740 0.614 BW 17x17 1.5 0.0470 1.37 0.0673 1.29 0.0745 0.937 0.0745 0.655 CE 14x14 1.8 0.0494 1.71 0.0705 1.64 0.0781 1.37 0.0781 1.37 CE 16x16 1.8 0.0489 1.71 0.0679 1.68 0.0724 1.52 0.0724 1.52 WE 14x14 1.9 0.0533 1.82 0.0725 1.76 0.0821 1.50 0.0821 1.50 WE 15x15 1.6 0.0475 1.35 0.0661 1.29 0.0746 0.859 0.0746 0.575 WE 17x17 1.6 0.0448 1.38 0.0646 1.26 0.0710 0.968 0.0710 0.691 NAC International 1.3-44
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 1.3-19 BWR Fuel Assembly Characteristics Fuel Class Characteristic 7X7 axa 9X9 1ox10 Base Fuel Typea SPC,GE SPC,GE SPC,GE SPC, GE, ABB Max Initial Enrichment (wt% 235LJ)b 4.5 4.5 4.5 4.5 Number of Fuel Rods 48 59 72 91d 49 60 74d 92d 61 76 96d,e 62 79 1Q0e 63 80 54c Max Assembly AveraQe Bumup (MWd/MTU) 60,000 60,000 60,000 60,000 Min Cool Time (years) 4 4 4 4 Min AveraQe Enrichment (wt% 235 U) 1.3 1.3 1.3 1.3 Max WeiQht (lb) per StoraQe Location See Note 1 See Note 1 See Note 1 See Note 1 Max Decay Heat (Watts) per Fuel Location 253 253 253 253 Each BWR fuel assembly may have a zirconium-based alloy channel :S 120 mil thick.
Assembly weight includes the weight of the channel.
Maximum initial enrichment is the peak planar-average enrichment.
Water rods may occupy more than one fuel lattice location. Fuel assembly to contain nominal number of water rods for the specific assembly design.
- All enrichment values are nominal pre-irradiation fabrication values.
- Spacers may be used to axially position fuel assemblies to facilitate handling.
Notes:
- 1. Maximum weight per storage location is 739 lbs (including fuel spacers) with a maximum contents weight of 62,656 lbs.
- Indicates assembly vendor/type referenced for fuel input data Fuel acceptability for loading is not restricted to the indicated vendor/type provided that the fuel assembly meets the limits listed in Table 6 .. 2.1-1. Table 6.2.1-2 contains vendor information by fuel rod array. Abbreviations are as follows:
General Electric/Global Nuclear Fuels (GE), Exxon/Advanced Nuclear Fuels/Siemens Power Corporation (SPC).
b Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235U for all basket/absorber combinations .
c May be composed of four subchannel clusters.
a Assemblies may contain partial-length fuel rods.
° Composed of four subchannel clusters.
NAC International 1.3-47
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Table 1.3-20 BWR Fuel Assembly Loading Criteria Geome:ry*,t 34 Min Min Max Max Number Number of Max Clad Clad Pellet Active Max Assembly of Fuel Partlal Length Pitch OD Thick. OD Length Loading Type Rods Rods 1 (inch) (inch) (Inch) (inch) (Inch) (MTU) 87 48A 48 NIA 0.7380 0.5700 0.03600 0.4900 144.0 0.1981 87 49A 49 NIA 0.7380 0.5630 0.03200 0.4880 146.0 0.2034 B7 49B 49 NIA 0.7380 0.5630 0.03200 0.4910 150.0 0.2037 88_59A 59 NIA 0.6400 0.4930 0.03400 0.4160 150.0 0.1828 88 60A 60 NIA 0.6417 0.4840 0.03150 0.4110 150.0 0.1815 88 608 60 NIA 0.6400 0.4830 0.03000 0.4140 150.0 0.1841 88_618 61 NIA 0.6400 0.4830 0.03000 0.4140 150.0 0.1872 88 62A 62 NIA 0.6417 0.4830 0.02900 0.4160 150.0 0.1921 88 63A 63 NIA 0.6420 0.4840 0.02725 0.4195 150.0 0.1985 88 64A 64 NIA 0.6420 0.4840 0.02725 0.4195 150.0 0.1996 88 6485 64 NIA 0.6090 0.4576 0.02900 0.3913 150.0 0.1755 89 72A 72 NIA 0.5720 0.4330 0.02600 0.3740 150.0 0.1803 89 74A 89 76A 89 79A 89 80A 810 91A 742 76 79 80 91 2 8
0.5720 0.5720 0.5720 0.5720 0.5100 0.4240 0.4170 0.4240 0.4230 0.3957 0.02390 0.02090 0.02390 0.02950 0.02385 0.3760 0.3750 0.3760 0.3565 0.3420 150.0 150.0 150.0 150.0 150.0 0.1873 0.1914 0.1979 0.1821 0.1906 810 92A 922 14 0.5100 0.4040 0.02600 0.3455 150.0 0.1946 810 96A5 962 12 0.4880 0.3780 0.02430 0.3224 150.0 0.1787 810_100A5 100 NIA 0.4880 0.3780 0.02430 0.3224 150.0 0.1861 1
Location of the partial length rods is illustrated in Figure 1.3-8.
2 Assemblies may contain partial-length fuel rods.
3 Assembly characteristics represent cold, unirradiated, nominal configurations.
4 Maximum channel thickness allowed is 120 mils (nominal).
5 Composed of four subchannel clusters.
NAC International 1.3-48
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 1.3-21 BWR Fuel Assembly Loading Criteria - Enrichment Limits for 87-Assembly and 82-Assembly Configurations with Axial Blanket Max. Initial Enrichments (wt% 235LJ)
Absorberb 0.027 109 Absorberb 0.0225 109 Absorber!> 0.02 10 8 g/cm 2 g/cm 2 g/cm2 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Basket Basket Basket Basket Basket Basket B7 48A 4.0% 4.5% 3.7% 4.5% 3.6% 4.4%
B7 49A 3.8% 4.5% 3.6% 4.4% 3.5% 4.3%
B7 49B 3.8% 4.5% 3.6% 4.4% 3.5% 4.2%
BB 59A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%
BB 60A 3.8% 4.5% 3.7% 4.4% 3.5% 4.2%
BB 60B 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
BB 618 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
BB 62A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%
BB 63A 3.B% 4.5% 3.6% 4.3% 3.4% 4.2%
BB 64A 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
BB 64B 3.6% 4.3% 3.4% 4.1% 3.3% 4.0%
B9 72A 3.8% 4.5% 3.6% 4.3% 3.4% 4.1%
B9 74A 3.7% 4.3% 3.4% 4.1% 3.4% 4.0%
B9 76A 3.5% 4.2% 3.4% 4.0% 3.3% 3.9%
B9 79A 3.7% 4.4% 3.4% 4.2% 3.3% 4.0%
B9 BOA 3.B% 4.5% 3.6% 4.3% 3.5% 4.2%
B10 91A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
B10 92A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%
B10 96A 3.7% 4.3% 3.5% 4.1% 3.4% 4.0%
B10_100A 3.6% 4.4% 3.5% 4.1% 3.4% 4.0%
Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235U for all basket/absorber combinations.
a Maximum planar average.
b Borated aluminum neutron absorber sheet effective areal 10B density .
- NAC International 1.3-49
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 1.3-22 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel Without Axial Blanket)
Max. Initial Enrichments (wt% 23 5ll)
Absorberb 0.027 10 8 g/cm2 Absorberb 0.0225 109 g/cm2 Absorberb 0.02 109 g/cm2 Fuel 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Type Basket Basket Basket Basket Basket Basket B7 48A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%
B7 49A 3.7% 4.5% 3.6% 4.3% 3.4% 4.1%
B7_49B 3.7% 4.5% 3.6% 4.3% 3.5% 4.2%
B8 59A 3.8% 4.5% 3.7% 4.4% 3.5% 4.3%
B8 60A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
B8 60B 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
B8 618 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
B8 62A 3.6% 4.4% 3.5% 4.2% 3.4% 4.1%
B8 63A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
B8 64A 3.7% 4.5% 3.5% 4.3% 3.4% 4.1%
B8 64B B9 72A B9 74A B9 76A 3.6%
3.7%
3.6%
3.5%
4.2%
4.4%
4.2%
4.1%
3.4%
3.5%
3.4%
3.3%
4.1%
4.2%
4.1%
4.0%
3.3%
3.4%
3.3%
3.2%
4.0%
4.1%
4.0%
3.8%
B9_79A 3.5% 4.2% 3.4% 4.1% 3.2% 3.9%
B9 80A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
B10 91A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
B10_92A 3.7% 4.4% 3.6% 4.2% 3.4% 4.1%
B10 96A 3.6% 4.2% 3.4% 4.1% 3.4% 4.0%
B10_100A 3.6% 4.3% 3.4% 4.0% 3.3% 3.9%
Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235U for all basket/absorber combinations.
- Maximum planar average.
Borated aluminum neutron absorber sheet effective areal 1°B density.
b NAC International 1.3-50
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Chapter 2 Structural Evaluation Table of Contents 2 STRUCTURAL EVALUATION ..................................................................................... 2-1 2.1 Description of Structural Design ............................................................................. 2.1-1 2.1.1 Discussion .................................................................................................... 2.1.1-1 2.1.2 Design Criteria ............................................................................................. 2.1.2-1 2.1.3 Weights and Centers of Gravity .................................................................. 2.1.3-l 2.1.4 Identification of Codes and Standards for Packaging ................................. 2.1.4-1 2.2 Materials .................................................................................................................. 2.2-1 2.2.1 Material Properties and Specifications ........................................................ 2.2.1-1 2.2.2 Chemical, Galvanic or Other Reactions ...................................................... 2.2.2-1 2.2.3 Effects of Radiation on Materials ............................................................... .2.2.3-1 2.3 Fabrication and Examination ................................................................................... 2.3-1 2.3.1 Fabrication ................................................................................................... 2.3.l-1 2.3.2 Examination ................................................................................................. 2.3.2-1 2.4 General Requirements for All Packages ................................................................. .2.4-1 2.4.1 Minimum Package Size .................................................................................. 2.4-1 2.4.2 Tamper-Indicating Feature ............................................................................. 2.4-1 2.4.3 Positive Closure .............................................................................................. 2.4-1 2.4.4 Chemical, Galvanic, Other Reactions and Radiation .................................... .2.4-1 2.4.5 Valves and Pressure Relief Devices ............................................................... 2.4-2 2.4.6 Loss or Dispersal of Radioactive Contents .................................................... 2.4-2 2.4.7 Surface Temperature During Transport .......................................................... 2.4-2 2.4.8 Continuous Venting During Transport .......................................................... .2.4-2 2.5 Lifting and Tie-Down Standards for All Packages ................................................ .2.5-1 2.5 .1 Lifting Devices ............................................................................................ 2.5 .1-1 2.5.2 Tie-Down Devices ....................................................................................... 2.5.2-1 2.6 Normal Conditions of Transport ............................................................................. 2.6-1 2.6.1 Heat .............................................................................................................. 2.6.l-1 2.6.2 Cold ............................................................................................................. 2.6.2-1 2.6.3 Reduced External Pressure .......................................................................... 2.6.3-1 2.6.4 Increased External Pressure ......................................................................... 2.6.4-1 2.6.5 Vibration ...................................................................................................... 2.6.5-1 2.6.6 Water Spray ................................................................................................. 2.6.6-1 2.6.7 Free Drop ..................................................................................................... 2.6.7-1 2.6.8 Corner Drop ................................................................................................. 2.6.8-1 2.6.9 Compression ................................................................................................ 2.6.9-1 2.6.10 Penetration ................................................................................................. 2.6.10-1 2.6.11 Fabrication Stresses ................................................................................... 2.6. l l-1 NAC International 2-i
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table of Contents (cont'd) 2.6.12 Transportable Storage Canister (TSC) Analysis -Normal Conditions of Transport ................................................................................................ 2.6.12-1 2.6.13 PWR Fuel Basket Analysis -Normal Conditions of Transport ............... .2.6.13-1 2.6.14 PWR DF Basket Analysis - Normal Conditions of Transport .................. 2.6.14-1 2.6.15 BWR Fuel Basket Analysis-Normal Conditions ofTransport ................ 2.6.15-1 2.6.16 GTCC Transportable Storage Canister and Waste Basket Liner Analysis - Normal Conditions of Transport .............................................. 2.6.16-1 2.6.17 Cask Cavity Spacer - Normal Conditions of Transport ............................. 2.6.17-1 2.7 Hypothetical Accident Conditions .......................................................................... 2.7-1 2.7.1 Free Drop (30-Foot) ..................................................................................... 2.7.1-1 2.7.2 Crush ............................................................................................................ 2.7.2-1 2.7.3 Puncture ........................................................................................................ 2.7.3-1 2.7.4 Structural Evaluation -Thermal (Fire Accident) ......................................... 2.7.4-1 2.7.5 Immersion-Fissile Material ....................................................................... 2.7.5-1 2.7.6 Immersion-All Packages ........................................................................... 2.7.6-1 2.7.7 Deep Water (290 psi) Immersion Test (for Type B Packages Containing more than 105 Ai) ......................................................................2.7.7-1 2.7.8 Transportable Storage Canister Analysis -Accident Conditions ................ 2.7.8-1 2.7.9 PWR Fuel Basket Analysis-Accident Conditions .................................... .2.7.9-1 *
- 2. 7 .10 PWR DF Basket Analysis - Accident Conditions .................................... .2. 7 .10-1 2.7.11 BWR Fuel Basket Analysis -Accident Conditions .................................. .2.7.11-1 2.7.12 GTCC-TSC and Waste Basket Liner Analysis -Accident Conditions .. .2.7.12-1 2.7.13 Fuel Basket Stability Evaluation ................................................................ 2.7.13-1 2.7.14 Cask Inner Shell Buckling Analysis-Accident Conditions ..................... 2.7.14-1
- 2. 7 .15 Cask Cavity Spacer - Hypothetical Accident Conditions ......................... .2. 7 .15-1
- 2. 7.16 Summary of Damage - Accident Conditions ............................................ .2. 7 .16-1 2.8 Accident Conditions for Air Transport of Plutonium .............................................. 2.8-1 2.9 Accident Conditions for Fissile Material Packages for Air Transport .................... 2.9-1 2.10 Special Form .......................................................................................................... 2.10-1 2.11 Fuel Rod Evaluations - Hypothetical Accident Conditions .................................. 2.11-1 2.11.1 PWR Fuel Rod Evaluation ......................................................................... 2.11.1-1 2.11.2 BWR Fuel Rod Evaluation ......................................................................... 2.1 l.2-l 2.11.3 RCCA Spacer Drop Evaluation ................................................................. 2.1 l .3-l 2.11.4 Side Drop Evaluation ................................................................................. 2.1 l .4-l 2.11.5 Thermal Evaluation of Fuel Rods .............................................................. 2.1 l.5-l 2.11.6 Fatigue Evaluation of Fuel Rods ............................................................... .2.1 l.6-l NAC International 2-ii
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 List of Tables (cont'd)
Table 2.7.11-1 BWR Fuel Tube Stresses, 30-ft Side Drop-0°, ksi ......................... 2.7.11-13 Table 2. 7.11-2 BWR Corner Weldment Plate Stresses, 30-ft Side Drop- 0°, ksi .... 2.7.11-13 Table 2.7.11-3 BWR Side Weldment Stresses, 30-ft Side Drop-0°, ksi ................. 2.7.11-13 Table 2.7.11-4 BWR Fuel Tube Stresses, 30-ft Side Drop-45°, ksi ....................... 2.7.11-14 Table 2.7.11-5 BWR Corner Weldment Plate Stresses, 30-ft Side Drop-45°,
ksi ...................................................................................................... 2.7.11-14 Table 2.7.11-6 BWR Side Weldment Stresses, 30-ft Side Drop- 45°, ksi ............... 2.7.11-14 Table 2.7.12-1 GTCC-TSC Pm Stresses -Internal Pressure (10 psig) ..................... 2.7.12.1-2 Table 2.7.12-2 GTCC-TSC Pm+ Pb Stresses -Internal Pressure (10 psig) ............. 2.7.12.1-2 Table 2.7.12-3 GTCC-TSC Critical Sections for the 30-Foot End-Drop Load Condition .......................................................................................... 2.7.12.2-2 Table 2.7.12-4 GTCC-TSC Pm Stresses Foot Top End Drop ........................... 2.7.12.2-3 Table 2.7.12-5 GTCC-TSC Pm+ Pb Stresses-30-Foot Top End Drop ................... 2.7.12.2-3 Table 2.7.12-6 GTCC-TSC Pm Stresses-30-Foot Top End Drop, Internal Pressure ............................................................................................. 2. 7 .12.2-4 Table 2.7.12-7 GTCC-TSC Pm + Pb Stresses - 30 Foot Top End Drop, Internal Pressure ............................................................................................. 2.7.12.2-4 Table 2.7.12-8 GTCC-TSC Pm Stresses Foot Bottom End Drop ..................... 2.7.12.2-5 Table 2.7.12-9 GTCC-TSC Pm+ Pb Stresses Foot Bottom End Drop ............ ,2.7.12.2-5 Table 2.7.12-10 GTCC-TSC Pm Stresses Foot Bottom End Drop, Internal Pressure ............................................................................................. 2.7.12.2-6 Table 2.7.12-11 GTCC-TSC Pm + Pb Stresses Foot Bottom End Drop, Internal Pressure ............................................................................... 2.7.12.2-6 Table 2.7.12-12 GTCC-TSC Critical Sections for the 30-Foot Side Drop Load Condition .......................................................................................... 2.7.12.3-2 Table 2.7.12-13 GTCC-TSC Pm Stresses- 30-Foot Side Drop .................................. 2.7.12.3-3 Table 2.7.12-14 GTCC-TSC Pm+ Pb Stresses- 30-Foot Side Drop .......................... 2.7.12.3-3 Table 2.7.12-15 GTCC-TSC Pm Stresses 30-Foot Side Drop, Internal Pressure ........ 2.7.12.3-4 Table 2.7.12-16 GTCC-TSC Pm+ Pb Stresses 30-Foot Side Drop, Internal Pressure.2.7.12.3-4 Table 2.7.12-17 GTCC-TSC Critical Sections for the 30-Foot Corner Drop Load Condition .......................................................................................... 2.7.12.4-2 Table 2.7.12-18 GTCC-TSC Pm Stresses Foot Top Comer Drop ...................... 2.7.12.4-3 Table 2.7.12-19 GTCC-TSC Pm+ Pb Stresses Foot Top Corner Drop ............... 2.7.12.4-3 Table 2.7.12-20 GTCC-TSC Pm Stresses 30-Foot Top Corner Drop, Internal Pressure ............................................................................................. 2. 7 .12.4-4 Table 2.7.12-21 GTCC-TSC Pm+ Pb Stresses 30-Foot Top Corner Drop, Internal Pressure ............................................................................................. 2. 7 .12.4-4 Table 2.7.12-22 GTCC-TSC Pm Stresses Foot Bottom Corner Drop ................. 2.7.12.4-5 Table 2.7.12-23 GTCC-TSC Pm+ Pb Stresses Foot Bottom Corner Drop ......... 2.7.12.4-5 Table 2.7.12-24 GTCC-TSC Pm Stresses 30-Foot Bottom Corner Drop, Internal Pressure ............................................................................................. 2. 7 .12.4-6 Table 2.7.12-25 GTCC-TSC Pm+ Pb Stresses 30-Foot Bottom Corner Drop, Internal Pressure ............................................................................... 2.7.12.4-6 Table 2.7.12-26 GTCC Waste Basket Liner Pm Stresses-30-Foot Drop Cases ....... 2.7.12.6-2 NAC International 2-xxiii
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Tables (cont'd)
Table 2.7.12-27 GTCC Waste Basket Liner Pm+Pb Stresses Foot Drop Cases ................................................................................................ 2.7.12.6-2 Table 2.7.13-1 Summary of Maximum Gap Changes at Pin-Slot Connections for PWR Basket ............................................................................. 2.7.13.2-21 Table 2.7.13-2 Summary of Maximum Gap Changes at Pin-Slot Connections for BWR Basket ............................................................................. 2.7.13.2-21 Table 2.7.14-1 Buckling Evaluation Load Cases and Results - Cask Inner Shell.. .... 2. 7.14-7 Table 2.7.14-2 Geometry Parameters for the MAGNATRAN Transport Cask .......... 2.7.14-8 Table 2.7.16-1 Critical Pm Stress Summary - Accident Conditions, ksi .................... 2. 7.16-3 Table 2.7.16-2 Critical Pm + Pb Stress Summary - Accident Conditions, ksi.. .......*... 2. 7.16-4 Table 2.11.3-1 Changes of Key Dimensions of Case 1 Model After Impact .............. 2.11.3-2 Table 2.11.3-2 Changes of Key Dimensions of Case 2 Model After Impact .............. 2.11.3-2 Table 2.11.6-1 Maximum Stress and Strain - PWR Fuel Rods Spectrum Analysis ... 2.11.6-2 Table 2.11.6-2 Maximum Stress and Strain - BWR Fuel Rods Spectrum Analysis .. 2.11.6-2 Table 2.12.2-1 Comparison of the MAGNATRAN and NAC-STC Cask and Impact Limiter Designs ..................................................................... 2.12.2-35 Table 2.12.2-2 Impact Limiter Benchmarking Analysis and Test Summary ............ 2.12.2-36 Table 2.12.2-3 Maximum Accelerations versus the Shallow Angle Drop for the NAC-STC CaskDesign .................................................................... 2.12.2-36 Table 2.12.2-4 Maximum Accelerations versus the Coefficient of Friction of the Impact Plane for Slapdown (5°) for the NAC-STC Cask Design ..... 2.12.2-36 Table 2.12.2-5 Parts Description of the NAC-STC-CY Cask Finite Element Side Drop Model ....................................................................................... 2.12.2-37 Table 2.12.2-6 Section Locations for Nonpuncture Cases ........................................ 2.12.2-79 Table 2.12.2-7 Section Stress Allowable Temperature Summary (Rounded Values) .............................................................................................. 2.12.2-80 Table 2.12.2-8 Section Locations for Side Puncture Model ..................................... 2.12.2-81 Table 2.12.2-9 Section Locations for Top Puncture Model ...................................... 2.12.2-82 Table 2.12.2-10 Section Locations for Bottom Puncture Model ................................ 2.12.2-82 NAC International 2-:xxiv
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 2.11 Fuel Rod Evaluations - Hypothetical Accident Conditions This section presents an evaluation of the PWR and BWR fuel rods for the 30-ft drop conditions for the MAGNATRAN system. Additionally, a thermal evaluation and a fatigue evaluation of the fuel rods are provided in Sections 2.11.5 and 2.11.6, respectively .
- NAC International 2.11-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 2.11.1 PWR Fuel Rod Evaluation This section presents the buckling evaluation for PWR fuel having cladding oxide layers that are 80 and 120 microns thick. A reduced cladding thickness is assumed due to the cladding oxide layer. The fuel assemblies are considered to be subjected to an axial loading with bounding acceleration corresponding to the 30-ft end drop condition. These analyses show that the maximum stresses in the PWR fuel rods remain below the yield strength in the design basis accident events and confirm that the fuel rods will return to their original configuration.
In the end drop orientation, the fuel rods are laterally restrained by the grids and come into contact with the fuel assembly base. The only vertical constraint for the fuel rod is the base of the assembly. Rather than evaluating a straight fuel assembly with all the grids present, the fuel assembly evaluated is considered to be bowed, and a fuel assembly grid may be missing (may still meet the acceptable configuration for undamaged fuel). The evaluation of the PWR fuel rods is based on the following representative samples.
Gap Between Fuel Cladding Cladding Fuel Rod Assembly and Fuel Diameter Thickness Pitch Tube Wall Fuel Assembly (in) (in) (in) (in)
WE 17x17 0.360 0.023 0.496 0.564 WE 15x15 0.422 0.024 0.563 0.561 WE 14x14 0.400 0.024 0.556 1.232 CE16x16 0.382 0.025 0.506 0.888 CE14x14 0.440 0.028 0.580 0.880 BW17x17 0.379 0.024 0.502 0.451 BW15x15 0.430 0.027 0.568 0.494 Review of the design basis fuel inventory indicates that the bounding fuel assembly is the WE 17x 17, as it has the lowest fuel rod cross-sectional moment of inertia. This fuel assembly is analyzed with an initial bow of0.01 inch (see NUREG-1864). The fuel assembly is analyzed for the following two cases:
Case 1. Intact grid Case 2. One grid damaged 2.11.1.1 Models for Cases 1 and 2 (Intact and Damaged Grid, No Gap)
A half-symmetry ANSYS model corresponding to a single row of fuel rods is used to model the case with intact and damaged grids. The ANSYS model for the 17x 17 assembly is shown in Figure 2.11.1-1. The fuel rod cladding is modeled with shell elements. Each grid is modeled
- NAC International 2.11.1-1
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 using brick elements to maintain the spacing between the fuel rods at the grid. The fuel tube is modeled using brick elements to restrict the lateral motion of the fuel assembly. Each of the fuel rods in the ANSYS model is simply supported at each end. Spring elements support the shell elements of the fuel rods at the locations of the grids and represent the fuel pellets. Static forces are applied between each grid to induce a maximum bow of 0.010 inch, which occurs between the lowest two grids. The purpose of the ANSYS model and solution is to provide the coordinates of the fuel clad for the LS-DYNA model. This is accomplished by obtaining a static solution with the ANSYS model, and then using the option to update the coordinates of the nodes based on the displacements from the solution. The LS-DYNA models are shown in Figure 2.11.1-2 and Figure 2.11.1-3 for Case 1 (intact grids) and Case 2 (one damaged grid),
respectively. A section of bottom grid in the intact grid model was removed to create the damaged grid model as shown in the Figure 2.11.1-3. In the analysis of fuel rod assemblies, the thickness of the cladding given above is reduced by 120 microns (0.0047 inch).
An initial velocity of 527 in/sec is defmed on all the nodes of fuel rod and fuel tube. The initial position of the fuel of the fuel assembly corresponds to the fuel assembly resting on the canister end. A deceleration curve, as shown in Figure 2.11.1-4, uses a maximum deceleration of 36 g's which bound the maximum axial acceleration in Table 2.6.7-37. The deceleration curve in Figure 2.11.1-4 is applied to the nodes of the elements representing the fuel tubes and fuel end fittings .
The deceleration time history is defined to result in a fmal velocity of 0 in/sec at the end of the 30-foot drop.
The LS-DYNA model employs the same nodes and elements as the ANSYS models (with the incorporation of the 0.010 inch bow). Elastic material properties are used in the ANSYS model and bilinear material properties are employed in the LS-DYNA model. Material properties for the zircaloy clad and the fuel pellet at 572°F (300°C) and 5/s strain rate are considered as shown in the following table.
Modulus of Yield Elasticity Density Strength (106 psi) (lb/in 3) (psi)
Fuel Clad 10.89(1) 0.237 90,625 Fuel Pellet (2) 0.396 (2)
<1l Based on 90% of the modulus at 150°C.
<2l Fuel pellet weight is considered and structural influence on fuel cladding stiffness is conservatively ignored.
NAC International 2.11.1-2
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 2.11.4 Side Drop Evaluation The basket side drop configuration is evaluated using a uniformly applied 60g's along the length of the basket. This bounds the accelerations developed in the transport cask side drop accident The analyzed bounding fuel rod length of 60.0 inches envelops all fuel types and includes the condition with a missing support grid in the fuel assembly. During a side drop, the maximum deflection of a fuel rod is based on the fuel rod spacing of the fuel assembly. Assuming a 17x 17 array (fuel assembly with the maximum number of rods), the maximum fuel rod deflection, including the 120-micron oxide layer, is:
(17-1) X (0.496-0.36+2x120Xl0-6x39.37) = 2.33 in.
The side drop loading is evaluated for three fuel rods, which corresponds to the limits of the stress modulus Z (ratio of the cross-sectional moment of inertia to the maximum radius to relate the maximum fiber stress (S) to the bending moment (M), S=M/Z ) and the maximum span, as shown in the following table.
Rod Diameter Clad Z (in 3) Span Case (inches) Thickness (10-3) (Inches)
(inches)
CE14x14 0.440 0.031 3.18 16.8 WE15x15 0.417 0.024 2.20 26.2 WE17x17 0.360 0.0205 1.33 20.6 ANSYS is used to perform a static analysis with a lateral loading of 60g. The model is shown in Figure 2.11 .4-1. The fuel rod is modeled with beam elements, and the properties for the fuel clad take into account the reduction of the outer radius by 0.0047 inch (120 microns). The density of the beam element material was based on the zircaloy clad (0.237 lb/in3) and the pellet density (0.396 lb/in3). The lateral constraints show the location of the grids used in the model, and the distance from the end of the fuel rod to the first support is 60 inches. The analyses confirm that the rod lateral displacement is 2.33 inches, which results when the fuel rod is assumed to be supported with a 60-inch distance between adjacent grids. Therefore, the location of the unsupported span along the fuel rod is not significant. The spacing for the adjacent grids is shown in the preceding table.
To represent the maximum gap of 2.33 inches, which the fuel rod can displace in the side drop, CONTAC52s were modeled at each node. The gap for each CONTAC52 was set to 2.33 inches to limit the lateral displacement of the fuel rod to 2.33 inches. The gap stiffness for each CONTAC52 was defined to be 106 lb/in, which simulates the resistance of the basket to the
- NAC International 2.11.4-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C lateral motion of the fuel rod. The lateral flexural stiffness of the fuel rod is considered to be insignificant compared to the stiffness of the basket. The effect of this stiffness, whether larger or smaller, would not influence the maximum stress. The maximum stress in the fuel rods is shown in the following table, and the allowable stress is the material yield strength at 752°F (69.6 ksi).
Case Maximum Stress (ksi) Factor of Safety CE14x14 37.1 +1.88 WE15x15 48.1 +1.45 WE17x17 46.3 +1.50 This confirms that the PWR fuel rods remain intact for a 60g side drop load condition.
Additional analyses are performed using the same fuel rod models for these PWR rods to incorporate the effect of DLF (Dynamic Load Factor). A maximum acceleration of 45.5g for cask side drop, as discussed in Section 2.6.7.5.1, is considered in the analyses. Since only the fuel clad is considered in the models, a factor of 1.25 is applied to the rod moment of inertia to implement the methodology to calculate cladding stress when using cladding-only properties as discussed in Section 2.3.3 ofNUREG-2224. A DLF of 1.75, which is the maximum response ratio corresponding to a half sine wave type impulse [Clough], is conservatively considered. The maximum stresses calculated by the finite element models and the amplified stresses with the conservative DLF are summarized in the following table. Using an allowable stress of the material yield strength at 752°F (69.6 ksi), the factors of safety are also provided. It is concluded that the PWR fuel rods remain structurally adequate for the side drop accident.
Case Maximum Stress Maximum Stress Factor of (ksi) With DLF (ksi) Safety CE14x14 29.2 51.1 1.36 WE15x15 35.5 62.1 1.12 WE17x17 34.5 60.4 1.15 The side drop evaluations for the PWR fuel rods in this section are bounding for the BWR fuel rods, since the (L/r) for the PWR fuel rod is significanly larger than the (L/r) for the BWR fuel rod, as discussed in Section 2.11.2. Therefore, no further evaluation of the BWR fuel rod is required.
NAC International 2.11.4-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 2.11.6 Fatigue Evaluation of Fuel Rods The section presents a fatigue evaluation for the PWR and BWR high burnup fuel assemblies for normal condition of transport for the MAGNATRAN system.
Three representative PWR fuel rods (CE 14x 14, WE 15 x 15 and WE 17x 17) and four BWR fuel rods (GE 7x7, GE 8x8, GE 9x9 and GE l0xl0) are considered in the evaluation. Two finite element models representing a single fuel rod for each of the PWR fuels are used (one without missing grid and the other with missing grids) to determine the stress and strain in the fuel cladding during normal conditions of transport. Similar models are used for the BWR fuel without missing grids. The fuel rod is modeled with ANSYS three-dimensional BEAM4 elements to represent the fuel clad only, and the properties for the fuel clad take into account the reduction of the outer radius by 0.0047 inch (120 microns). The density of the clad is adjusted to account for the mass of the fuel pellet. The locations of the grids are modeled as simply supports in the lateral directions. The model for the missing grids case for PWR fuel rods has a maximum span of 60 inches.
Response spectrum analyses are performed for the fuel rods using response spectra of the transport cask platform from seven test cases as documented in the ENSA/DOE rail cask test
[SAND2018-13258R]. The response spectra include acceleration data in the axial and two lateral directions of the fuel rods up to 1,000 Hz frequency. Missing mass and Close Mode Grouping options are considered in the analyses.
The maximum stress and strain of the fuel rods from the spectrum analyses for the PWR fuel and BWR fuel are summarized in Table 2.11.6-1 and Table 2.11.6-2, respectively. The maximum strain is 0.046% for the PWR fuel and 0.043% for the BWR fuel, which are well below the 0.06% end point of the Lower-Bound Fatigue Curve as shown in Table 2-5 and Figure 2-12 of NUREG-2224, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel. Therefore, fatigue is not a concern of the high bumup PWR and BWR fuel assemblies for transport conditions .
- NAC International 2.11.6-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 2.11.6-1 Maximum Stress and Strain - PWR Fuel Rods Spectrum Analysis Fuel Type Missing Grid Max. Stress (ksi) Max. Strain(%)
CE 14x14 No 4.76 0.044 Yes 4.87 0.045 WE 15x15 No 3.93 0.036 Yes 3.95 0.036 WE 17x17 No 4.98 0.046 Yes 4.12 0.038 Table 2.11.6-2 Maximum Stress and Strain - BWR Fuel Rods Spectrum Analysis Fuel Type Max. Stress (ksi) Max. Strain (%)
GE7x7 4.68 0.043 GE 8x8 4.19 0.038 GE 9x9 4.09 0.038 GE 10x10 3.86 0.035 NAC International 2.11.6-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- 68. Tietz, T.E., "Determination of the Mechanical Properties of High Purity Lead and a 0.05%
Copper-Lead Alloy," Stanford Research Institute, Menlo Park, CA, W ADC Technical Report 57-695, ASTIA Document Number 151165, April 1958.
- 69. Weaver, William Jr. and James M. Gere, "Matrix Analysis of Framed Structures," Second Edition, D. Van Nostrand Company, New York, 1980.
- 70. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 2001 Edition with 2003 Addenda.
- 71. Machinery's Handbook," 25 th Edition, Industrial Press, New York, 1996.
- 72. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendix F, Rules For Evaluation of Service Loadings With Level D Service Limits," American society of Mechanical Engineers, New York, New York, 2001 Edition with 2003 Addenda.
- 73. Blake, Alexander, Practical Stress Analysis in Engineering Design, 2nd Ed., Marcel Dekker, Inc., 1990.
- 74. Boyer, H.E., "Atlas of Stress-Strain Curves," ASM International, Metals Park, OH, 1987.
- 75. Field Manual of the Interchange Rules as Adopted by the Association of American Railroads, Rule 88, "Mechanical Requirements for Acceptance," Washington, DC, 1986.
- 76 .
77.
ASME Boiler and Pressure Vessel Code, Code Cases - Nuclear Components, Code Case N-595-4, Requirements for Spent Fuel Storage Canisters Section ill, Division l,"
Approved May 12, 2004.
NUREG-2224, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel (Draft Report for Comment), Office of Nuclear Material Safety and Safeguards, July, 2018.
- 78. SAND2018-13258R, Data Analysis ofENSA/DOE Rail Cask Tests, Spent Fuel and Waste Disposition, US Department of Energy, Spent Fuel and Waste Science and Technology, November 19, 2018 .
- NAC International 2.12.1-5
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 4 CONTAINMENT The MAGNATRAN transport cask containment boundary is designed and analyzed to ensure the containment of the cask contents in accordance with 10 CFR 71 (71.43 and 71.51). The containment boundary is tested to ANSI NI 4.5-1997 leaktight criteria and is designed, fabricated and inspected in accordance with ASME Code,Section III, Subsection NB, with the exception of code stamping. The cask is designed to facilitate leakage testing of the containment boundary penetrations (i.e., lid and lid port cover) prior to transport to confirm the containment boundary.
The transportable storage canister (TSC), while not a component of the cask containment system, is evaluated for maximum pressure under normal and a HAC conditions. The sealed TSC represents the expected transport configuration, and bounds any other potential TSC condition as the sealed TSC contains a high pressure helium backfill that can, and is, considered to be released into the cask cavity. No credit is taken for the TSC as a pressure boundary under any transport condition.
Criticality evaluation were performed for conditions both crediting and not crediting the TSC sealed boundary for moderator exclusion from the fissile material region. In the context of moderator exclusion, the TSC is credited with serving the 10 CFR 71.55(c) function of being a special design feature that prevents a single packing error from permitting leakage into the fissile material region. Leakage testing of the cask containment seals assures that the containment does not leak. Regardless of credit applied to the TSC confinement boundary to prevent water in-leakage, the containment function is retained by the transport cask body .
- NAC International 4-1
MAGNATRAN Transport Cask SAR December 2020
- Docket No. 71-9356 Table 5.8-2 List of Tables (cont'd)
PWR Fuel Assembly Nonzirconium Alloy-Based Hardware Mass ........ 5.8.1-3 Revision 20C Table 5.8-3 PWR Sample In-Core Characteristics ...................................................... 5.8.1-3 Table 5.8-4 BWR Fuel Assembly Geometry Data ...................................................... 5.8.1-4 Table 5.8-5 BWR Fuel Assembly Nonzirconium Alloy-Based Hardware Quantities ................................................................................................. 5. 8.1-4 Table 5.8-6 BWR Sample In-Core Characteristics ..................................................... 5.8.1-4 Table 5.8-7 Response Method to Direct Calculation Comparison - Transport Cask. 5.8.2-8 Table 5.8-8 Sample Gamma Response Calculation for Normal Condition Radial Surface-Fuel Centerline (35 GWd/MTU, 2.3 wt%, 7.6-Year Cooled 16a Hybrid) .............................................................................................. 5.8.2-9 Table 5.8-9 Sample Neutron Response Calculation for Normal Condition Radial Surface-Fuel Centerline (35 GWd/MTU, 2.3 wt%, 7.6-Year Cooled 16a Hybrid) ............................................................................................ 5.8.2-10 Table 5.8-10 Sample Hardware Gamma (Upper End-Fitting) Response Calculation for Normal Condition Radial Surface - Upper End-Fitting Elevation (35 GWd/MTU, 2.3 wt%, 7.6-Year Cooled 17a Hybrid) ...................... 5.8.2-11 Table 5.8-11 PWR Fuel Region Homogenization Sample Calculation ...................... 5.8.3-20 Table 5.8-12 PWR Nonfuel Hardware Homogenization Sample Calculation ............ 5.8.3-20 Table 5.8-13 Key PWR Basket Geometry Features .................................................... 5.8.3-20 Table 5.8-14 PWR Minimum Cool Time Solution .......................................... ,.......... 5.8.3-21 Table 5.8-14b Low Burnup PWR Fuel Loading Table ................................................. 5.8.3-21 Table 5.8-15 Loading Table for PWR Fuel (23 kW/Cask) ......................................... 5.8.3-22 Table 5.8-16 Loading Table for PWR Fuel (21.85 kW/Cask) .................................... 5.8.3-24 Table 5 .8-17 PWR Normal Conditions Maximum Dose Rate Summary ................... 5.8.3-25 Table 5 .8-18 PWR Hypothetical Accident Event Maximum Dose Rate Summary .... 5.8.3-25 Table 5 .8-19 BWR Fuel Region Homogenization Sample Calculation ...................... 5.8.4-18 Table 5.8-20 BWR Nonfuel Hardware Homogenization Sample Calculation ........... 5.8.4-18 Table 5.8-21 BWR Fuel Region Homogenized Material Description ........................ 5.8.4-19 Table 5.8-22 Key BWR Basket Geometry Features ................................................... 5.8.4-20 Table 5.8-23 BWR Minimum Cool Time Solution ..................................................... 5.8.4-20 Table 5.8-23b Low Burnup BWR Fuel Loading Table ................................................. 5.8.4-20 Table 5.8-24a Loading Table for BWR Fuel (22 kW/Cask) ......................................... 5.8.4-21 Table 5.8-24b Loading Table for BWR Fuel (20.9 kW/Cask) ...................................... 5.8.4-23 Table 5.8-25 BWR Normal Conditions Maximum Dose Rate Summary ................... 5.8.4-25 Table 5.8-26 BWR Hypothetical Accident Event Maximum Dose Rate Summary ... 5.8.4-25 Table 5.8-27 Sample Core Type BPRA Hardware Summary - Westinghouse 15x15 Core ............................................................................................... 5.8.5-8 Table 5.8-28 Bounding Regional Nonfuel Hardware Masses ....................................... 5.8.5-8 Table 5.8-29 Allowed BPRA Exposure and Cool Time Combinations ........................ 5.8.5-9 Table 5.8-30 BPRA Maximum Transport Detector Surface Dose Rates ...................... 5.8.5-9 Table 5.8-31 Allowed GTPD Exposure and Cool Time Combinations ........................ 5.8.5-9 Table 5.8-32 Thimble Plug Maximum Surface Dose Rates ........................................ 5.8.5-10 Table 5.8-33 Westinghouse Fuel Reduced 2-Meter Dose Rates to Load NFHW ....... 5.8.5-10 NAC International 5-lx
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Table 5.8-34 List of Tables (cont'd)
Additional Fuel Assembly Cool Time Required to Load NFHW ........ 5.8.5-10 Revision 0 Table 5.8-35 HFRA vs. BPRA Source Comparison .................................................. 5 .8.5-11 Table 5.8-36 Bounding CEA Mass Quantities Descriptions ........................................ 5.8.6-5 Table 5.8-37 CEA Maximum Surface Dose Rates ....................................................... 5.8.6-5 Table 5.8-38 Gamma Source Comparison for CEA Primary Absorber Materials ....... 5.8.6-6 Table 5.8-39 Dose Rates from Ag-In-Cd Based CEAs as a Function of Exposure and Cool Time ......................................................................................... 5.8.6-7 Table 5.8-40 CEA Maximum Exposure and Minimum Cool Time Summary ............ 5 .8.6-7 Table 5.8-41 [DELETED] ............................................................................................. 5.8.8-1 Table 5.8-42 [DELETED] ............................................................................................. 5 .8.8-1 Table 5.8-43 [DELETED] ............................................................................................. 5 .8.8-1 Table 5.8-44 [DELETED] ............................................................................................. 5 .8.8-1 Table 5.8-45 Zoned Fuel and Profile Effects on Source Magnitudes .......................... 5.8.9-4 Table 5.8-46 Millstone Zoned Fuel Effects on Source Magnitudes ............................. 5.8.9-5 Table 5.8-47 Damaged Fuel Material Summary- 14a PWR Fuel .......................... 5.8.10-73 Table 5.8-48 PWR Damaged Fuel Comparison - Normal Conditions 2 Meter Radial Dose Rates ............................................................................... 5.8.10-74 Table 5.8-49 Additional Cool Time Required for 23 kW Damaged PWR Fuel Table 5.8-50 Table 5.8-51 Table 5.8-52 Table 5.8-53 Table 5.8-54 Contents .............................................................................................. 5.8.10-74 Normal Condition Maximum Damaged PWR Fuel Dose Rates ........ 5.8.10-75 Accident Event Maximum Damaged PWR Fuel Dose Rates ............. 5.8.10-75 GTCC Source Spectrum ..................................................................... 5. 8.11-17 GTCC Material Description ................................................................ 5 .8.11-17 Normal Condition Maximum GTCC Waste Dose Rates .................... 5.8.11-18 Table 5.8-55 Accident Condition Maximum GTCC Waste Dose Rates .................. 5.8.11-18 Table 5.8-56 Radial Surface Tally Detector Results for WE 14x14 Fuel Assembly 5.8.12-20 Table 5.8-57 Comparison of Surface Dose Rates with Localized Peaking .............. 5.8.12-20 Table 5.8-58 Low Bumup PWR Fuel Loading Table-22 kW/Cask ......................... 5.8.14-1 Table 5.8-59 Loading Table for PWR Fuel (22 kW/Cask) ......................................... 5.8.14-2 Table 5.8-60 Loading Table for PWR Fuel (20.9 kW/Cask) ...................................... 5.8.14-4 Table 5.8-61 Additional Cool Time to Load Non-Fuel Hardware (Reduced Heat Load- 22kW PWR - Configuration) ............................ 5.8.14-5 NAC International 5-x
MAGNATRAN Transport Cask SAR December 2020
- Docket No. 71-9356 5 SHIELDING EVALUATION The MAGNATRAN transport cask meets the 10 CFR 71 requirements for exclusive use dose Revision 20C rate limits. The optimized multiwall design of the transport cask provides an efficient shielding arrangement for the transportation of a TSC containing up to 37 undamaged PWR fuel assemblies in the 37 PWR basket assembly or up to 87 undamaged BWR fuel assemblies in the 87 BWR basket assembly. The cask is also designed to transport a TSC containing up to four damaged fuel cans (DFCs) in the DF Basket Assembly. The DF Basket Assembly has a capacity ofup to 37 undamaged PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations. Each DFC may contain an undamaged PWR fuel assembly, a damaged PWR fuel assembly, or PWR fuel debris equivalent to one PWR fuel assembly. Undamaged PWR fuel assemblies may be placed directly in the DFC locations of a DF Basket Assembly. The cask is also designed to transport a TSC containing up to 32,000 pounds of GTCC waste in a GTCC waste basket liner. The transport cask is assigned a Transport Index, maximum dose rate in mrem/hr at lm from the package, of 28.4 (TI= 28.4) for undamaged fuel, 20.9 (TI= 20.9) for damaged fuel, and 2.4 (TI=2.4) for GTCC waste based on the requirement of 10 CFR 71.4 and the analysis presented in this chapter. The TI for damaged fuel is below that for undamaged fuel as bounding undamaged fuel radial dose rates are obtained from BWR and "long" TSC PWR payloads. Damaged fuel is restricted to the "short" TSC. A spacer above the "short TSC will retain source below the region of minimum radial cask shielding.
The shielding design criteria for the transport cask are in accordance with the requirements established in 10 CFR Part 71.47 for normal conditions of transport and 10 CFR Part 71.51 for hypothetical accident events. The 10 CFR 71.4 7 requirements for the exclusive use transport of spent fuel under normal conditions of transport include the following:
- The dose rate on the surface of the enclosed package must not exceed 1,000 mrem/hr.
- The dose rate on the outer surfaces of the transport vehicle must not exceed 200 mrem/hr.
- The dose rate on a plane 2 meters from the lateral surfaces of the vehicle must not exceed 10 mrem/hr (NAC analysis limit 2-meter dose rates to 9 .5 mrem/hr).
- The dose rate in any normally occupied positions of the vehicle must not exceed 2 mrem/hr.
The IO CFR 71.51 requirements state that the dose rate under hypothetical accident events must not exceed 1 rem/hr at I meter from the surface of the transport cask.
The transport cask with its impact limiters is securely attached to the bed of a conveyance vehicle during transport. To restrict unauthorized personnel from gaining access to the transport
- cask during transport, a personnel barrier is installed around the transport cask. The personnel NAC International 5-1
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 barrier consists of a metal frame structure covered with expanded metal (i.e., a metal grating or screen) and is securely attached to the bed of the conveyance vehicle during normal transportation. Thus, the loaded transport cask, with the personnel barrier on the conveyance vehicle ready for transport, meets the exclusive use definition of a closed conveyance.
MAGNATRAN is designed as a single-length transport cask that will hold a TSC of various lengths. A top spacer is used for the short TSCs. TSCs may be closed with either an all stainless steel closure lid or a composite carbon steel and stainless steel lid assembly. BWR evaluations and PWR evaluations are performed with the composite closure lid assembly. The composite lid assembly bounds the all stainless steel lid in shielding evaluations due to the lower density of carbon steel.
This chapter describes the shielding design and the analysis used to establish bounding radiological dose rates for the transport of PWR and BWR fuel and GTCC waste. PWR fuel assemblies may contain non-fuel hardware (control) components as described in Chapter 1.
Evaluated components are control element assemblies (RCCA/CEAs) or rod control cluster assemblies (RCCAs), burnable poison rod assemblies (BPRAs), thimble plugs (also referred to as flow mixers or guide tube plugging device [GTPD]), primary and secondary neutron sources, and hafnium flux reduction assemblies (HFRA). Fuel assemblies may contain up to five
- irradiated stainless steel rods in place of fuel rods. Transport of undamaged and damaged PWR fuel assemblies is permitted. Loading of damaged fuel is limited to short canisters requiring a cask cavity spacer. This configuration spaces damaged fuel away from the gap between the neutron shield and the upper impact limiter gap. BWR assemblies are limited to undamaged fuel.
Minimum cool times prior to fuel transport are specified as a function of minimum assembly average fuel enrichment and maximum assembly average bumup (MWd/MTU). To minimize the number of loading tables, PWR and BWR fuel assemblies are grouped by bounding fuel and hardware mass. Key characteristics of each assembly grouping are shown in Section 5.2. Refer to Section 5.8.3 for loading tables used to generate bounding system dose rates for the PWR system and Section 5.8.4 for the BWR system. Adjustments, as necessary, to PWR fuel assembly minimum cool time to account for PWR damaged fuel inclusion, non-fuel hardware insertion, and loading of irradiated stainless steel rods within the assembly are included in the relevant subsections to Section 5.8. Uncertainties in the SAS2H-generated heat loads for high bumup (> 45 GWd/MTU assembly average) fuel assemblies are accounted for in the loading tables by derating the system heat load by 5%.
Source terms for the various vendor-supplied fuel types are generated using the SCALE 4.4 sequence as discussed in Section 5.2. Three-dimensional MCNP shielding evaluations provide NAC International 5-2
MAGNATRAN Transport Cask SAR April 2019
- Docket No. 71-9356 Normal Operating Conditions Revision 0 The maximum radial and axial dose rates calculated for the package under normal conditions of transport for damaged PWR fuel are shown in Table 5.1-5. The locations of the maximum dose rates during normal conditions of transport relative to the transport cask body for damaged PWR fuel are shown in Figure 5.1-3. Additional bounding dose rate information is provided in Section 5.6.4.
Hypothetical Accident Events Table 5 .1-6 provides the calculated hypothetical accident condition dose rates. The locations of the maximum hypothetical accident dose rates relative to the transport cask body are shown in Figure 5.1-4. Additional bounding dose rate information is provided in Section 5.6.4.
The bounding PWR fuel type, cool time, bumup and initial enrichment for each detector under normal conditions and accident events are given in Table 5.1-10.
5.1.2.3 Greater than Class C (GTCC) Waste GTCC waste may be transported in the MAGNATRAN canister. Only the "short" TSC is evaluated for GTCC contents. This configuration contains a top spacer in the transport cask cavity separating the GTCC source region from the impact limiter to neutron shield assembly gap where maximum dose rates occur. The shielding evaluation of GTCC waste is based on quantities of waste from the decommissioning of Zion Nuclear Power Station (ZNPS).
Bounding source terms were generated based on the maximum GTCC weight loaded in the TSC, as defined by project-specific operational control (32,000 lbs). Maximum dose rates for the normal condition of transport for GTCC waste are presented in Table 5.1-7 and Figure 5.1-5.
The maximum dose rates for the accident condition of transport for GTCC waste are presented in Table 5.1-8 and Figure 5.1-6. The Transport Index for GTCC waste shipments is 2.4 mrem/hr, while the maximum heat load is 1.7 kW .
- NAC International 5.1-5
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Figure 5.1-1 Location of Maximum Dose Rates for Normal Conditions ofTransport-Undamaged Fuel Figure 5.1-2 Location of Maximum Dose Rates for Hypothetical Accident Events-Undamaged Fuel NAC International 5.1-6
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 Table 5.1-1 Key Transport Cask Shielding Features Table 5.1-2 Key TSC Shielding Features Table 5.1-3 Normal Conditions Maximum Total Dose Rate Summary-Undamaged Fuel
- NAC International 5.1-9
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.1-4 Maximum Hypothetical Accident Event Dose Rate Summary -
Undamaged Fuel Table 5.1-5 Normal Conditions Maximum Total Dose Rate Summary-Damaged PWR Fuel Table 5.1-6 Maximum Hypothetical Accident Event Dose Rate Summary -
Damaged PWR Fuel NAC International 5.1-10
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Figure 5.3-1 Enveloping Axial Burnup Profile for PWR Fuel 12~---------------------------------,
11 I 080 eo C.
9 m
'0 08
.§ go O7 z
06 0 547 0 547 05 I* - - Normalized Bumup - Envelope I 0 10 20 30 40 50 60 70 80 90 JOO Distance From Bottom of Active Fuel (% Active Core Height)
- Figure 5.3-2 Enveloping Axial Burn up Profile for BWR Fuel 14~---------------------------------,
I 22,.__ _ _~--~~-------to: 122 I2 I 18 C.
E m 08
'0
~
.; 06 E
0 z
04 02 0 043 0
0 JO 20 30 40 50 60 70 80 90 JOO Distance From Bottom of Active Fuel (3/4 Active Core Height) 1- -----Max Normalized Bum up - Envelope I
- NAC International 5.3-3
"NAC PROPRIETARY INFORMATION REMOVED" MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Figure 5.3-3 Average and Midplane Normal Condition Surface Dose Rates as a Function of Burn up (Fixed Heat Load)
NAC International 5.3-4
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision O Table 5.6-1 Sample Minimum Cool Time Calculation for WE 17 Fuel Assembly (Years)
Minimum Cool Time (years)
Minimum Nonnal Enrichment Decay Heat 2m+Railcar Active
[wt% 2351.J] (23 kW/cask) (8.5 mrem/hr)
- Limiting Constraint 2.7 19.9 19.7 20.0 Decay Heat 2.9 19.5 17.4 19.6 Decay Heat 3.1 19.4 15.3 19.4 Decay Heat 3.3 19.1 13.3 19.1 Decay Heat 3.5 18.8 11.5 18.8 Decay Heat 3.7 18.6 9.9 18.7 Decay Heat 3.9 18.4 8.7 18.4 Decay Heat 4.1 18.2 7.8 18.3 Decay Heat 4.3 18.1 7.1 18.1 Decay Heat 4.5 17.9 6.6 18.0 Decay Heat 4.7 17.7 6.2 17.7 Decay Heat 4.9 17.6 5.9 17.6 Decay Heat
- Table 5.6-2 ANSI Standard Neutron Flux-To-Dose Rate Factors Enemv (MeV) 2.5E-08 1.0E-07 1.0E-06 (rem/hr)/(n/cm2/sec) 3.67E-06 3.67E-06 4.46E-06 1.0E-05 4.54E-06 1.0E-04 4.18E-06 1.0E-03 3.76E-06 1.0E-02 3.56E-06 1.0E-01 2.17E-05 5.0E-01 9.26E-05 1.0 1.32E-04 2.5 1.25E-04 5.0 1.56E-04 7.0 1.47E-04 10.0 1.47E-04 14.0 2.08E-04 20.0 2.27E-04
- 10 CFR 71 dose rate limit is 10 mrem/hr at 2 meters. NAC conservatively reduces the limit to 9.5 mrem/hr. The
- limit for WE 17xl 7 fuel (no NFHW) is reduced further to 8.5 mrem/hr to allow loading any non-fuel hardware while remaining under 9.5 mrem/hr at 2 meters.
NAC International 5.6-25
MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.6-3 ANSI Standard Gamma Flux-To-Dose Rate Factors Energy (MeV) (rem/hr)/(y/cm 2/sec) Energy (MeV) (rem/hr)/(y/cm 2/sec) 0.01 3.96E-06 1.4 2.51E-06 0.03 5.82E-07 1.8 2.99E-06 0.05 2.90E-07 2.2 3.42E-06 0.07 2.58E-07 2.6 3.82E-06 0.1 2.83E-07 2.8 4.01E-06 0.15 3.79E-07 3.25 4.41 E-06 0.2 5.01E-07 3.75 4.83E-06 0.25 - 6.31E-07 4.25 5.23E-06 0.3 7.59E-07 4.75 5.60E-06 0.35 8.78E-07 5 5.80E-06 0.4 9.85E-07 5.25 6.01E-06 0.45 1.08E-06 5.75 6.37E-06 0.5 1.17E-06 6.25 6.74E-06 0.55 1.27E-06 6.75 7.11E-06 0.6 1.36E-06 7.5 7.66E-06 0.65 1.44E-06 9 8.77E-06 0.7 1.52E-06 11 1.03E-05 0.8 1.68E-06 13 1.18E-05 1 1.98E-06 15 1.33E-05 Table 5.6-4 Bounding Source Type for Transport Cask Shielding - Undamaged Fuel Detector Fuel Burnup Enrichment Cool Time Model Biasing Type [GWd/MTUJ [wt% 235 UJ [Years]
Normal Radial 09b 30 4.3 7.7 Conditions Top Axial 14b 60 3.3 37.6 Bottom Axial 14a 60 3.3 32.5 Accident Radial 07a 60 3.3 42.3 Events Top Axial 14b 60 3.3 37.6 Bottom Axial 14a 60 3.3 32.5 NAC International 5.6-26
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C energy lines by example. Energy lines used in the dose rate assessment are shown in italics and represent 99.8+% of the total dose rate (99.9+% of gamma and 99.8+% of neutron dose rates).
Energy lines with no (0) source magnitude are not included in the tables.
The response function method allows the MCNP weight window (acceleration) map to be optimized for a particular source energy, producing an increased number of particles scoring per source particle. The response function also allows for a significant reduction in the number of MCNP shielding runs, thereby increasing the number of particles per MCNP run (based on fixed computer resources). For example, a single PWR fuel type has approximately 20,000 source runs associated with it, requiring the same number of MCNP runs to determine a complete dose rate set. The same dose rate set for fuel gamma and neutron cases may be generated using approximately 20 MCNP runs (one per relevant neutron and gamma energy line) using the response function.
The applicability of the response function method to determine dose rates at the range of bumups requested is based on the ability to apply fresh fuel material composition based dose rate responses to spent fuel. A single dose rate response may be generated for all bumups of a particular assembly type, as dose rates have historically been calculated using a fresh fuel material composition (see Table 5.5-4). To confirm the accuracy of this assumption, radial surface dose rates are calculated for a sample high bumup (PWR 60 GW d/MTU) fuel assembly in the normal condition and accident event models and compared to the fresh fuel results. This calculation is done with subcritical multiplication within MCNP turned off ("nonu" card is used).
Dose rate profiles for fresh and spent fuel isotopics are shown in Figure 5.8-7 and Figure 5.8-8 and demonstrate the acceptability of the fresh fuel assumption (i.e., there is no significant dose rate change associated with the fresh fuel model). Fissile material content changes will effect subcritical multiplication and therefore neutron and n-y dose rates. Calculation determining dose rates for comparison to 10 CFR 71 limits apply a conservative fissile material content (5 wt%
fresh fuel for undamaged PWR and BWR systems, and 2 wt°/4 for the PWR damaged fuel system). Further discussion on the effect of fissile material content is included in the damaged fuel analysis section (Section 5.8.10) .
- NAC International 5.8.2-3
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Figure 5.8-1 Comparison of Response Method to Direct Solution: Normal Condition Radial Surface - PWR Low Burn up/Low Enrichment Figure 5.8-2 Comparison of Response Method to Direct Solution: Normal Condition Radial Surface - PWR Medium Burnup/Medium Enrichment NAC International 5.8.2-4
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 5.8.4 87-Assembly BWR System This section presents the detailed evaluation of the transport cask loaded with BWR fuel assemblies.
5.8.4.1 BWR Fuel and Basket Models The three-dimensional shielding evaluation includes a homogenized fuel assembly model and a detailed three-dimensional basket model.
Fuel Assembly Model Based on the fuel assembly physical parameters provided in Table 5.8-4 and the hardware masses in Table 5.8-5, homogenized treatments of fuel assembly source regions are developed.
The homogenized fuel assembly is represented in the model as a stack of boxes with width equal to the fuel assembly width. The height of each box corresponds to the modeled height of the corresponding assembly region.
Sample homogenizations for the source regions for the 09b assembly are shown in Table 5.8-19 and Table 5.8-20. The resulting fuel compositions on an atom/barn-cm basis are shown in Table 5.8-21. Similar compositions sets are generated for the remaining fuel assembly hybrids .
Note that the zirconium fuel assembly channel is not included in the model.
Basket Model The basket is composed of coated carbon steel tubes, pinned together at the corners, and held together by side and corner weldments. Forty-five fuel tubes, in combination with the weldments, form 89 fuel openings. Two openings are located below the TSC port covers. To minimize exposure and meet ALARA constraints, the basket capacity is reduced to 87 assemblies by prohibiting fuel assembly loading into the openings beneath the ports. Pin spacers or tube extensions maintain the tube axial spacing within the TSC cavity. The neutron absorber sheets are not modeled. In dry transport, the presence of the neutron absorber sheets provides minimal shielding and could, therefore, be removed without a significant increase in exposure.
Key basket characteristics are shown in Table 5.8-22. Radial and axial sketches of the BWR basket within the TSC are shown in Figure 5.8-22 and Figure 5.8-23, respectively.
5.8.4.2 Minimum Cool-time Specification SAS2H generates heat loads for all BWR fuel types listed in Section 5.8.1.2. Based on a 22 kW per cask (252.8 W per assembly) heat load and transport dose rate limits, minimum allowed cool times for each fuel type are calculated. Calculated heat loads and radiation sources account for
- fuel material (actinide and fission product) and hardware (light element) generated sources .
NAC International 5.8.4-1
MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Minimum cool times are conservatively rounded up to the nearest one-tenth of a year. A sample minimum cool-time calculation for a BWR assembly is shown in Table 5.8-23. Allowed low burnup (up to 30,000 MWd/MTU) fuel loadings are shown in Table 5.8-23b. Note that the listed minimum cool times at each burnup step are bounding for all BWR fuel types and initial enrichments above the minimum enrichment specified. Collapsing the fuel type and initial enrichment-dependent minimum cool time matrix to a single value may result in a minimum cool time longer than individual values presented for higher burnups in the detailed tables that follow.
The resulting minimum cool times are listed in an assembly specific loading table (see Tables 5.8-24a and 5.8-24b).
Note that the loading table removes combinations of high burnup and low enrichment (e.g.,
45 GWd/MTU and 1.9 wt% 235U) from the payload definition. Source term data covering these combinations is generated, but produces unrealistic source terms due to the complete consumption of fissile uranium early in the burn up cycle and the SAS2H input of a fixed power density. To maintain power density, ORJGEN-S (SAS2H) will substantially increase flux levels, which would not occur during core operation of the assembly, to produce fissile material and to produce power by nonthermal fission. The increased flux level "breeds" higher actinides, which in tum increase source significantly. Because a high bumup and low enrichment combination would require repeated reinsertion of a burned assembly, the combination is excluded.
5.8.4.3 Transport Cask Dose Rates Using the dose rate response method, transport cask dose rates are tabulated for all allowed cool time, bumup, and initial enrichment combinations for each of the assembly types. Minimum cool times are based on the shortest cool time meeting both heat load and dose rate limits (rounded up to the nearest tenth of a year). Resulting normal conditions of operation maximum dose rates as a function of distance from the transport cask surface are shown in Figure 5.8-24 for the cask radial surface, Figure 5.8-27 for the cask top, and Figure 5.8-28 for the cask bottom.
The radial I foot dose rate results correspond to the personnel barrier. Breakdowns of the cask radial surface and two-meter dose rates into the source components are shown in Figure 5.8-25and Figure 5.8-26, respectively.
The azimuthal dose distribution over the potential streaming paths of the lower trunnion rotation pockets and neutron shield compartment walls is illustrated in Figure 5.8-29. Surface and I foot azimuthal detectors at the axial height of the expansion foam in the neutron shield are shown in Figure 5.8-29B and Figure 5.8-29C, respectivley. There is discemable peaking at the cask surface due to the neutron shield top expansion foam, but the peak flattens by the I foot boundary corresponding to the personnel barrier. Maximum surface and I foot dose rates occur NAC International 5.8.4-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C higher along the cask at the gap between the top impact limiter and the neutron shield assemblies.
The dose rate profile at the exposed cask surface below the top impact limiter and above the neutron shield assemblies is shown in Figure 5.8-30 and is the location of maximum cask surface dose rates. Cask accident event dose rates are plotted in Figure 5.8-31 for the cask radial surface at one meter and Figure 5.8-32 and Figure 5.8-33 for the cask top and bottom axial surfaces, respectively. Assembly maximum dose rate summaries for normal and accident events are listed in Table 5.8-25 and Table 5.8-26, respectively.
The bounding payloads for each detector surface are listed as follows.
Initial Bumup Enrichment Cool Time Model Direction Fuel Type (GWd/MTU) (wt% 235U) (vrs)
Normal Radial 09b 30 4.3 7.7 Top 10a 45 2.7 21.5 Bottom 09a 60 3.3 38.2 Accident Radial 07a 60 3.3 42.3 Top 10a 45 2.7 21.5 Bottom 09a 60 3.3 38.2 Note that the limiting fuel type, cool time, bumup, and initial enrichment for normal conditions radial dose rates is based on the limiting surface dose rate because the 9 .5 mrem/hr normal conditions 2 meter radial dose rate limit occurs for numerous fuel assembly/source term combinations .
- NAC International 5.8.4-3
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision O Figure 5.8-22 BWR Basket Top View
+ +
+
+ +
+
S!daWeldment
+ + + + + +
+ + + + + .625
.26
+ + + + + +
+ +
Tubo TSC Shd
+ +
+
+
+
I
.250 Woll I
I a - S t o o l O* Noutron , _ , _ . \
0
- Cabon Stool ~
- Fuel "-nbly
.100 ' :,-...._ ____ ,
Note: Neutron absorber sheets are not modeled in the BWR shielding evaluation.
NAC International 5.8.4-4
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Figure 5.8-27 BWR Bounding Top Axial Dose Rate Profiles for Normal Conditions of Transport
- NAC International 5.8.4-9
"NAC PROPRIETARY INFORMATION REMOVED" MAGNA TRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Figure 5.8-28 BWR Bounding Bottom Axial Dose Rate Profiles for Normal Conditions of Transport NAC International 5.8.4-10
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020
- Docket No. 71-9356 Figure 5.8-31 Revision 20C BWR Bounding Radial lm Dose Rate Profile for Accident Events of Transport
- NAC International 5.8.4-15
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 Figure 5.8-32 BWR Bounding Top Axial Dose Rate Profiles for Accident Events of Transport NAC International 5.8.4-16
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020
- Docket No. 71-9356 Figure 5.8-33 Revision 20C BWR Bounding Bottom Axial Dose Rate Profiles for Accident Events of Transport
- NAC International 5.8.4-17
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 Table 5.8-19 BWR Fuel Region Homogenization Sample Calculation Area Area Volume Fraction of Components Component [cm2] Fraction U02 Void Clad Fuel 5.6593E+01 2.8809E-01 2.8809E-01 Gap 2.8033E+00 1.4271E-02 1.4271E-02 Clad 1.8455E+01 9.3945E-02 9.3945E-02 Water Rods 4.6792E-01 2.3820E-03 2.3820E-03 Inside Water Rods 1.5024E+00 7.6484E-03 7.6484E-03 Interstitial* 1.1662E+02 5.9366E-01 5.9366E-01 Total 1.9644E+02 1.0000E+00 2.8809E-01 6.1558E-01 9.6327E-02 Table 5.8-20 BWR Nonfuel Hardware Homogenization Sample Calculation Assy 55 Modeled Dimensions Mass 55 Volume Height Volume Volume Region [k!1fassy] [cm 3/assy] [cm] [cm 3/assy] Fraction Lower Nozzle 4.74 5.9698E+02 17.1704 3.3730E+03 1.7699E-01 Lower Plenum t 0.00 0.0000E+00 1.5875 3.1185E+02 0.0000E+00 Fuel Hardware 0.25 3.1033E+01 381.0000 7.4844E+04 4.1463E-04 Upper Plenum 1.71 2.1480E+02 28.6421 5.6265E+03 3.8177E-02 Upper Nozzle 2.08 2.6196E+02 19.0500 3.7422E+03 7.0003E-02
- Space in fuel assembly width envelope outside fuel rods and water rod(s).
t Represents fuel rod end-cap.
NAC International 5.8.4-18
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision O Table 5.8-21 BWR Fuel Region Homogenized Material Description Density Density Material fa/cm 3l Nuclide/Element fatoms/b-cml Lower End-Fitting 1.4053 CHROMIUM 3.0924E-03 MANGANESE 3.0808E-04 IRON 1.0532E-02 NICKEL 1.3698E-03 Lower Plenum 2.5998 CHROMIUM 3.0110E-05 TIN 1.9782E-04 IRON 3.5042E-05 NITROGEN 5.5902E-05 ZIRCONIUM 1.6857E-02 Active Fuel 3.6315 URANIUM-235 3.3873E-04 URANIUM-238 6.3545E-03 ZIRCONIUM 4.0975E-03 CHROMIUM 7.3187E-06 TIN 4.8083E-05 NITROGEN 1.3587E-05 OXYGEN 1.3383E-02 IRON 8.5174E-06 Upper Plenum* 0.3830 CHROMIUM 6.6797E-04 TIN 6.0773E-06 MANGANESE 6.6453E-05 IRON 2.2728E-03 NITROGEN 1.7174E-06 NICKEL 2.9546E-04 ZIRCONIUM 5.1787E-04 Upper End-Fitting 0.5558 CHROMIUM 1.2231 E-03 MANGANESE 1.2185E-04 IRON 4.1655E-03 NICKEL 5.4177 E-04
- Contains a zirconium alloy component due to the addition of the fuel rod end-plug material in the homogenization.
NAC International 5.8.4-19
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-22 Key BWR Basket Geometry Features Table 5.8-23 BWR Minimum Cool Time Solution Minimum Cool Time (: *ears)
Minimum Nonnal Enrichment Decay Heat 2m+Railcar Active
[wt% 235 U] (22 kW/cask) (9.5 mrem/hr) Limiting Constraint 2.7 19.3 22.7 22.7 Dose Rate -
2.9 19.0 20.0 20.0 Dose Rate 3.1 18.7 17.8 18.8 Decay Heat 3.3 3.5 3.7 3.9 4.1 4.3 18.5 18.3 18.0 17.9 17.7 16.9 16.0 15.2 14.5 13.8 18.6 18.3 18.0 17.9 17.7 Decay Heat ..
Decay Heat Decay Heat _
Decay Heat Decay Heat 17.6 13.2 17.7 Decay Heat _
4.5 17.4 12.6 17.5 Decay Heat 4.7 17.3 12.1 17.4 Decay Heat 4.9 17.2 11.6 17.3 Decay Heat Note: Sample calculation for the 09b fuel type at 45 GWd/MTU. Cask surface (normal), lm (accident), and axial (normal and accident) dose rates are significantly below limits at all cool-times meeting assembly heat load limits and are therefore not listed.
Table 5.8-23b Low Burn up BWR Fuel Loading Table Min. Assembly Avg. Minimum Initial Enrichment Cool Time wt°/c, 23 ears 10 000 1.3 6.3 15 000 1.5 8.6 20 000 1.7 10.3 25 000 1.9 11.9 30,000 2.1 13.7 NAC International 5.8.4-20
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-24a Loading Table for BWR Fuel (22 kW/Cask)
Minimum Initial Assembly Average Bum up~ 30 GWd/MTU Assembly Avg. Minimum Coolina Time {vears)
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 6.5 12.3 5.8 13.7 5.3 13.0 13.5 2.3 s E < 2.5 6.3 11.6 5.7 13.0 5.2 12.3 12.8 2.5 s E < 2.7 6.3 11.0 5.7 12.3 5.1 11.7 12.2 2.7 s E < 2.9 6.2 10.3 5.6 11.8 5.1 11.1 11.6 2.9 s E < 3.1 6.1 9.8 5.6 11.2 5.0 10.5 11.1 3.1 s E < 3.3 6.0 9.3 5.5 10.7 5.0 10.0 10.6 3.3 s E < 3.5 6.0 8.8 5.5 10.2 4.9 9.6 10.0 3.5 s E < 3.7 6.0 8.4 5.4 9.8 4.9 9.1 9.6 3.7 s E < 3.9 5.9 8.0 5.4 9.4 4.9 8.8 9.2 3.9 s E < 4.1 5.9 7.7 5.3 9.0 4.8 8.4 8.9 4.1 s E < 4.3 5.9 7.4 5.3 8.7 4.8 8.0 8.5 4.3 s E < 4.5 5.8 7.0 5.3 8.4 4.8 7.7 8.2 4.5 s E < 4.7 5.8 6.8 5.2 8.1 4.7 7.5 7.9 4.7 s E <4.9 5.8 6.5 5.2 7.8 4.7 7.2 7.6 E ~ 4.9 5.7 6.3 5.2 7.6 4.7 6.9 7.4 Minimum Initial 30 < Assembly Average Bumup ~ 35 GWd/MTU Assembly Avg. Minimum Coollna Time (vears)
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235l.J (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 - - - - - - -
2.3 s E < 2.5 8.9 14.3 7.6 15.6 6.6 15.0 /,)
15.5 2.5 s E < 2.7 8.8 13.5 7.5 14.8 6.5 14.1 14.6 2.7 s E < 2.9 8.6 12.7 7.3 14.0 6.4 13.4 13.9 2.9 s E < 3.1 8.5 12.0 7.2 13.4 6.3 12.7 13.2 3.1 s E < 3.3 8.4 11.4 7.2 12.8 6.3 12.1 12.6 3.3 s E < 3.5 8.3 10.8 7.1 12.2 6.2 11.5 12.0 3.5 s E < 3.7 8.2 10.3 7.0 11.7 6.1 11.0 11.5 3.7 s E < 3.9 8.1 9.8 6.9 11.2 6.0 10.6 11.0 3.9 s E < 4.1 8.0 9.3 6.9 10.8 6.0 10.1 10.6 4.1 s E < 4.3 8.0 8.9 6.9 10.4 6.0 9.7 10.1 4.3 s E < 4.5 8.0 8.7 6.8 10.0 6.0 9.3 9.8 4.5 s E < 4.7 7.9 8.6 6.8 9.6 5.9 8.9 9.4 4.7 s E < 4.9 7.8 8.6 6.7 9.3 5.9 8.6 9.1 E ~ 4.9 7.8 8.6 6.7 9.0 5.9 8.3 8.8
- NAC International 5.8.4-21
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-24a Loading Table for BWR Fuel (22 kW/Cask) (cont'd)
Minimum Initial 35 < Assembly Average Bumup ~ 40 GWd/MTU Assembly Avg. Minimum Cooling Time (years)
Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235U (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 ~ E < 2.3 - - - - - - -
2.3 s E < 2.5 - - - - - - -
2.5 s E < 2.7 14.6 16.9 12.2 18.0 10.0 17.3 17.7 2.7 s E < 2.9 13.3 15.8 10.7 17.0 8.7 16.3 16.7 2.9 s E < 3.1 13.1 14.9 10.5 16.0 8.5 15.4 15.8 3.1 s E < 3.3 12.9 14.1 10.3 15.2 8.4 14.6 15.0 3.3 ~ E < 3.5 12.6 13.9 10.1 14.5 8.3 13.8 14.3 3.5 s E < 3.7 12.5 13.7 10.0 13.8 8.2 13.2 13.6 3.7 ~ E < 3.9 12.4 13.6 9.9 13.3 8.0 12.6 13.0 3.9sE<4.1 12.2 13.5 9.8 12.7 8.0 12.1 12.5 4.1 ~ E < 4.3 12.0 13.4 9.7 12.3 7.9 12.0 11.9 4.3 s E < 4.5 11.9 13.3 9.6 12.2 7.9 12.0 11.5 4.5 ~ E < 4.7 11.9 13.1 9.6 12.1 7.8 11.9 11.2 4.7 s E < 4.9 11.8 13.0 9.5 12.0 7.8 11.8 11.2 E~4.9 11.8 13.0 9.4 12.0 7.7 11.8 11.1 Minimum Initial Assembly Avg.
Enrichment wt% 235U (E) 2.1 ~ E < 2.3 BWR/2-3 7x7 7x7 40 < Assembly Average Bumup ~ 45 GWd/MTU BWR/4-6 Minimum Cooling Time (years)
BWR/2-3 BWR/4-6 BWR/2-3 8x8 8x8 9x9 BWR/4-6 9x9 BWR/4-6 10x10 2.3 s E < 2.5 - - - - - - -
2.5 ~ E < 2.7 - - - - - - -
2.7 s E < 2.9 22.3 24.0 19.9 22.9 17.4 22.7 21.5 2.9 s E < 3.1 19.7 21.4 17.2 20.3 14.8 20.0 19.4 3.1 ~ E < 3.3 18.9 20.5 15.4 19.1 12.3 18.8 18.2 3.3 ~ E< 3.5 18.7 20.2 15.2 18.8 11.9 18.6 17.4 3.5 s E < 3.7 18.5 20.0 15.0 18.7 11.7 18.3 17.2 3.7 ~ E < 3.9 18.2 19.9 14.7 18.5 11.5 18.0 17.1 3.9 s E < 4.1 18.1 19.6 14.6 18.2 11.4 17.9 16.9 4.1 ~ E < 4.3 17.8 19.5 14.3 18.1 11.3 17.7 16.7 4.3 s E < 4.5 17.8 19.4 14.3 18.0 11.2 17.7 16.5 4.5 s E < 4.7 17.6 19.2 14.1 17.8 11.1 17.5 16.5 4.7 ~ E < 4.9 17.4 19.0 14.0 17.8 11.0 17.4 16.3 E2:'.4.9 17.3 18.9 13.8 17.7 10.9 17.3 16.2 NAC International 5.8.4-22
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-24b Loading Table for BWR Fuel (20.9 kW/Cask)
- NAC lnternatlonal 5.8.4-23
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-24b Loading Table for BWR Fuel (20.9 kW/Cask) (cont'd)
NAC International 5.8.4-24
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 5.8-25 BWR Normal Conditions Maximum Dose Rate Summary Table 5.8-26 BWR Hypothetical Accident Event Maximum Dose Rate Summary
- NAC International 5.8.4-25
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Chapter 6 Criticality Evaluation Table of Contents 6 CRITICALITY EVALUATION ..................................................................................... 6-1 6.1 Description of Criticality Design ............................................................................... 6.1.1-1 6.1.1 Design Features .............................................................................................. 6.1.1-1 6.1.2 Summary Criticality Evaluation Results ........................................................ 6.1.2-1 6.1.3 Criticality Safety Index .................................................................................. 6.1.3-1 6.2 Material Contents ....................................................................................................... 6.2.1-1 6.2.1 Fissile Material ............................................................................................... 6.2.1-1 6.2.2 Nonfuel Hardware .......................................................................................... 6.2.2-1 6.2.3 Greater Than Class C (GTCC) Material ........................................................ 6.2.3-1 6.3 General Considerations .............................................................................................. 6.3 .1-1 6.3.1 Model Configuration ...................................................................................... 6.3.1-1 6.3.2 Material Properties ......................................................................................... 6.3.2-1 6.3.3 Computer Codes and Cross-Section Libraries ............................................... 6.3.3-1 6.3.4 Demonstration of Maximum Reactivity ........................................................ 6.3.4-1 6.4 Single Package Evaluation (10 CFR 71.55) ................................................................. 6.4-1 6.4.1 Configuration/Discussion .............................................................................. 6.4.1-1 6.4.2 Results ............................................................................................................ 6.4.2-1 6.5 Evaluation of Package Arrays Under Normal Conditions of Transport (10 CFR 71.59) .......................................................................................................... 6.5.1-1 6.5.1 Configuration ................................................................................................. 6.5.1-1 6.5.2 Results ............................................................................................................ 6.5.2-1 6.6 Package Arrays Under Hypothetical Accident Conditions (10 CFR 71.59) ............. 6.6.1-1 6.6.1 Configuration ................................................................................................. 6.6.1-1 6.6.2 Fuel Loading Configuration for Maximum Reactivity .................................. 6.6.2-1 6.6.3 Results ............................................................................................................ 6.6.3-1 6.7 Fissile Material Packages For Air Transport ................................................................ 6.7-1 6.8 Benchmark Evaluations ............................................................................................. 6.8.1-1 6.8.1 Criticality Code Validation ............................................................................ 6.8.1-1 6.8.2 Depletion Code Sequence Validation ............................................................ 6.8.2-1 6.9 References ..................................................................................................................... 6.9-1 6.10 Criticality Evaluation Detail .................................................................................... 6.10.1-1 6.10.1 PWR Undamaged Fuel Basket Evaluations ................................................. 6.10.1-1 6.10.2 PWR Damaged Fuel Criticality Evaluation ................................................. 6.10.2-1 6.10.3 BWR Undamaged Fuel Criticality Evaluation ............................................ 6.10.3-1 6.10.4 Moderator Exclusion Criticality Evaluation ................................................ 6.10.4-1
- NAC International 6-i
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 List of Figures Figure 6.1.1-1 82-Assembly BWR Basket Configuration ............................................... 6.1.1-3 Figure 6.1.1-2 PWR Damaged Fuel Basket Configuration ............................................. 6.1.1-4 Figure 6.2.1-1 BWR Partial Length Fuel Rod Location Sketches .................................. 6.2.1-2 Figure 6.3.1-1 Generic Tube Cross-Section and Developed Cell Model Sketches ......... 6.3.1-5 Figure 6.3.1-2 TSC Model Sketch ................................................................................... 6.3.1-6 Figure 6.3.1-3 Transport Cask Model Sketch .................................................................. 6.3 .1-7 Figure 6.8.1-1 LEU USLSTATS Output for EALCF (MCNP5 v 1.3 / ENDF/B Vl) ... 6.8.1-10 Figure 6.8.1-2 LEU keffversus Fuel Enrichment (MCNP5 v 1.3 / ENDF/B VI) .......... 6.8.1-12 Figure 6.8.1-3 LEU keffversus Rod Pitch (MCNP5 v 1.3 / ENDF/B VI) ..................... 6.8.1-12 Figure 6.8.1-4 LEU keff versus Fuel Pellet Diameter (MCNP5 v 1.3 / ENDF /B VI) .... 6.8.1-13 Figure 6.8.1-5 LEU keffversus Fuel Rod Outside Diameter (MCNP5 v 1.3 /
ENDFIB VI) ........................................................................................... 6.8.1-13 Figure 6.8.1-6 LEU keffversus Hydrogen/235U Atom Ratio (MCNP5 v 1.3 /
ENDF/B VI) ........................................................................................... 6.8.1-14 Figure 6.8.1-7 LEU keffversus Soluble Boron Concentration (MCNP5 v 1.3 /
ENDF/B VI) ........................................................................................... 6.8.1-14 Figure 6.8.1-8 LEU keff versus Cluster Gap Thickness (MCNP5 v 1.3 /
ENDF/B VI) ........................................................................................... 6.8.1-15 Figure 6.8.1-9 LEU keff versus 1OB Plate Loading (MCNP5 v 1.3 / ENDF /B VI) ....... 6.8.1-15 Figure 6.8.1-10 LEU keff versus Energy of Average Neutron Lethargy Causing Fission (MCNP5 v 1.3 / ENDF/B VI) ................................................... 6.8.1-16 Figure 6.8.1-11 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-12 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-13 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-14 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-15 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-16 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-17 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-18 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-19 [DELETED] ........................................................................................... 6.8.1-17 Figure 6.8.1-20 Spent Fuel USLSTATS Output for EALCF .......................................... 6.8.1-18 Figure 6.8.1-21 Spent Fuel keffvs. Energy of Average Neutron Lethargy Causing Fission .................................................................................................... 6.8.1-21 Figure 6.8.1-22 [DELETED] ........................................................................................... 6.8.1-21 Figure 6.8.1-23 Spent Fuel keffvs. Water-to-Fuel Volume Ratio .................................... 6.8.1-22 234 8 Figure 6.8.1-24 Spent Fuel keffVS. U/23 U Ratio********************************************************* 6.8.1-22 Figure 6.8.1-25 Spent Fuel keffvs. 235 U/2 38U Ratio ......................................................... 6.8.1-23 Figure 6.8.1-26 Spent Fuel keff vs. 238Pu/2 38U Ratio ........................................................ 6.8.1-23 Figure 6.8.1-27 Spent Fuel keff vs. 239Pu/2 38 U Ratio ........................................................ 6.8.1-24 Figure 6.8.1-28 Spent Fuel keff vs. 240Pu/2 38 U Ratio ........................................................ 6.8.1-24 NAC International 6-ii
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision O List of Figures (cont'd)
Figure 6.8.1-29 Spent Fuel keff vs. 241 Pu/238 U Ratio ........................................................ 6.8.1-25 Figure 6.8.1-30 Spent Fuel kerrvs. 242 Pu/238 U Ratio ........................................................ 6.8.1-25 Figure 6.8.1-31 Spent Fuel keffvs. 241 Am/2 38 U Ratio ...................................................... 6.8.1-26 Figure 6.8.1-32 LEU Fresh Fuel keffversus Fuel Enrichment for MCNP6.l .................. 6.8.1-26 Figure 6.8.1-33 LEU Fresh Fuel ketr versus Rod Pitch for MCNP6.l ............................. 6.8.1-27 Figure 6.8.1-34 LEU Fresh Fuel keff versus Fuel Pellet Diameter for MCNP6. l ........... 6.8.1-27 Figure 6.8.1-35 LEU Fresh Fuel keffversus Fuel Rod Outside Diameter for MCNP6.l 6.8.1-28 Figure 6.8.1-36 LEU Fresh Fuel keff versus Hydrogen/2 35 U Atom Ratio for MCNP6. l. 6.8.1-28 Figure 6.8.1-37 LEU Fresh Fuel keffversus Soluble Boron Concentration for MCNP6.l ............................................................................................... 6.8.l-29 Figure 6.8.1-38 LEU Fresh keff versus Cluster Gap Thickness for MCNP6.l ................ 6.8.1-29 Figure 6.8.1-39 LEU Fresh Fuel keff versus 1°B Plate Loading for MCNP6.l ................ 6.8.1-30 Figure 6.8.1-40 LEU Fresh Fuel keffversus Energy of Average Neutron Lethargy Causing Fission for MCNP6. l ............................................................... 6.8.1-30 Figure 6.8.2-1 [DELETED] ............................................................................................. 6.8.2-2 Figure 6.8.2-2 [DELETED] ............................................................................................. 6.8.2-2 Figure 6.8.2-3 [DELETED] ............................................................................................. 6.8.2-2 Figure 6.8.2-4 [DELETED] ............................................................................................. 6.8.2-2 Figure 6.10.1-1 PWRFuel Tube .................................................................................... 6.10.1-27 Figure 6.10.1-2 PWR Basket Structure ......................................................................... 6.10.1-28 Figure 6.10.1-3 MCNP Analysis - Undamaged PWR Basket Assembly Basket ................................................................................................... 6.10.1-29 Figure 6.10.1-4 Nominal Tube Interface Width and Nominal Tube Outer Width ........ 6.10.1-54 Figure 6.10.1-5 Nominal Tube Interface Width and Minimum Tube Outer Width ...... 6.10.1-54 Figure 6.10.1-6 Minimum Tube Interface Width and Maximum Tube Outer Width ... 6.10.1-55 Figure 6.10.1-7 Minimum Tube Interface Width and Minimum Tube Outer Width .... 6.10.1-55 Figure 6.10.1-8 TRITON Actinide and Fission Product Isotopics Generation Sample Input ..................................................................................................... 6.10.1-56 Figure 6.10.1-9 Bounding Loading Curve Generation- CE14-0.036 g/cm2 1°B ....... 6.10.1-59 Figure 6.10.1-10 Bounding Loading Curve Generation - CEl 6 - 0.036 g/cm2 1°B ....... 6.10.1-59 Figure 6.10.1-11 Bounding Loading Curve Generation- WE14-0.036 g/cm2 10B ...... 6.10.1-60 Figure 6.10.1-12 Bounding Loading Curve Generation - WE15 - 0.036 g/cm2 1°B ...... 6.10.1-60 Figure 6.10.1-13 Bounding Loading Curve Generation- WEI 7- 0.036 g/cm2 10B ...... 6.10.1-61 Figure 6.10.1-14 Bounding Loading Curve Generation - BWl 5 - 0.036 g/cm2 1°B ...... 6.10.1-61 Figure 6.10.1-15 Bounding Loading Curve Generation -BWl 7 - 0.036 g/cm2 1°B ...... 6.10.1-62 Figure 6.10.1-16 [DELETED] ......................................................................................... 6.10.1-62 Figure 6.10.1-17 Schematic of Optional PWR Basket Configurations ........................... 6.10.1-63 Figure 6.10.1-18 RCCA Unit Cell and 1/4 Assembly Model for TRITON/NEWT Depletion .............................................................................................. 6.10.1-64 Figure 6.10.1-19 RCCA Allowed Basket Location (Zone A) ......................................... 6.10.1-65 NAC International 6-iii
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Figures (cont'd)
Figure 6.10.1-20 RCCA Depletion Input (TRITON) ...................................................... 6.10.1-66 Figure 6.10.1-21 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-22 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-23 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-24 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-25 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-26 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-27 [DELETED] ......................................................................................... 6.10.1-78 Figure 6.10.1-28 Undamaged PWR Basket Moderator Density Study ........................... 6.10.1-79 Figure 6.10.1-29 Reactivity vs. Burnup for WE 15x15 Fuel Assembly ......................... 6.10.1-79 Figure 6.10.1-30 Reactivity vs. Enrichment for WE 15x15 Fuel Assembly ................... 6.10.1-80 Figure 6.10.1-31 keff Distribution for RCCA (Ag-In-Cd) Benchmarking ....................... 6.10.1-80 Figure 6.10.1-32 Histogram of Reactivity Results for Additional Insert Bands in Fuel Depletion Analysis ............................................................................... 6.10.l-81 Figure 6.10.1-33 Histogram of Reactivity Results for Mixed-Mode Depletion in Fuel Depletion Analysis ............................................................................... 6.10.1-82 Figure 6.10.2-1 PWR Damaged Fuel Basket Sketch ....................................................... 6.10.2-8 Figure 6.10.2-2 Damaged Fuel Basket Moderator Density Study ................................... 6.10.2-9 Figure 6.10.3-1 BWR Fuel Tubes .................................................................................. 6.10.3-10 Figure 6.10.3-2 BWR Basket Structure ......................................................................... 6.10.3-11 Figure 6.10.3-3 BWR 87-Assembly Basket Optional Neutron Absorber Sheet Locations .............................................................................................. 6.10.3-12 Figure 6.10.3-4 BWR 82-Assembly Basket Optional Neutron Absorber Sheet Locations .............................................................................................. 6.10.3-13 Figure 6.10.3-5 MCNP Transport Cask Model -BWR 87-Assembly Basket.. ............ 6.10.3-14 Figure 6.10.3-6 MCNP Transport Cask Model -BWR 82-Assembly Basket.. ............ 6.10.3-38 Figure 6.10.3-7 Maximum Tube Interface Width & Maximum Tube Corner Grind .... 6.10.3-56 Figure 6.10.3-8 Maximum Tube Interface Width & Minimum Tube Corner Grind ..... 6.10.3-56 Figure 6.10.3-9 Minimum Tube Interface Width & Maximum Tube Corner Grind ..... 6.10.3-57 Figure 6.10.3-10 Minimum Tube Interface Width & Minimum Tube Corner Grind ..... 6.10.3-57 Figure 6.10.3-11 Maximum Reactivity Moderation Graphical Results .......................... 6.10.3-58 Figure 6.10.4-1 Homogeneous UO2 Sphere - VISED XZ Slice ..................................... 6.10.4-3 NAC International 6-iv
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Tables Table 6.1.1-1 Effective Areal Density as a Function of Absorber Credit.. .................... 6.1.1-5 Table 6.1.2-1 PWR Fuel Assembly Loading Criteria (Assembly Description) .............................................................................................. 6.1.2-3 Table 6.1.2-2 BWR Fuel Assembly Loading Criteria (Assembly Description) ............ 6.1.2-4 Table 6.1.2-3 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel with Axial Blanket) ......................................................... 6.1.2-5 Table 6.1.2-4 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel without Axial Blanket) ................................................... 6.1.2-6 Table 6.1.2-5 [DELETED] ............................................................................................. 6.1.2-6 Table 6.2.1-1 PWR Fuel Assembly Characteristics ....................................................... 6.2.1-3 Table 6.2.1-2 BWR Fuel Assembly Characteristics ....................................................... 6.2.1-4 Table 6.3.2-1 Fuel Assembly Material Densities and Compositions ............................. 6.3.2-2 Table 6.3.2-2 Basket, TSC and Cask Material Densities and Compositions ................. 6.3.2-3 Table 6.4.2-1 [DELETED] ............................................................................................. 6.4.2-2 Table 6.5.2-1 [DELETED] ............................................................................................. 6.5.2-2 Table 6.6.3-1 [DELETED] ............................................................................................. 6.6.3-4 Table 6.6.3-2 [DELETED] ............................................................................................. 6.6.3-4 Table 6.6.3-3 [DELETED] ............................................................................................. 6.6.3-4
- Table 6.8.1-1 Table 6.8.1-2 Table 6.8.1-3 LEU Range of Applicability for Complete Set of 186 Benchmark Experiments ........................................................................................... 6.8.1-31 MCNP5 vl.3 LEU Correlation Coefficients and USLs for Benchmark Experiments ........................................................................................... 6.8.1-31 LEU MCNP Validation Statistics .......................................................... 6.8.1-32 Table 6.8.1-4 [DELETED] ........................................................................................... 6.8.1-53 Table 6.8.1-5 [DELETED] ........................................................................................... 6.8.1-53 Table 6.8.1-6 [DELETED] ........................................................................................... 6.8.1-53 Table 6.8.1-7 Spent Fuel Range of Applicability for Complete Set of 215 Benchmark Experiments ........................................................................ 6.8.1-54 Table 6.8.1-8 Spent Fuel Correlation Coefficients and USLs for Benchmark Experiments ........................................................................................... 6.8.1-55 Table 6.8.1-9 Spent Fuel MCNP Validation Statistics ................................................. 6.8.1-56
- NAC International 6-v
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 List of Tables (cont'd)
Table 6.8.1-10 LEU Fresh Fuel Correlation Coefficients and USLs for Benchmark Experiments for MCNP6.l .................................................................... 6.8.1-66 Table 6.8.1-11 Set ofNuclides for Actinide-Only Bumup Credit.. ............................... 6.8.1-66 Table 6.8.1-12 Set of Additional Nuclides for Actinide and Fission Product Bumup Credit. ..................................................................................................... 6.8.1-66 Table 6.8.1-13 LEU Fresh Fuel MCNP6.l Validation Statistics ................................... 6.8.1-67 Table 6.8.2-1 [DELETED] ............................................................................................. 6.8.2-2 Table 6.8.2-2 [DELETED] ............................................................................................. 6.8.2-2 Table 6.8.2-3 [DELETED] ............................................................................................. 6.8.2-2 Table 6.8.2-4 [DELETED] ............................................................................................. 6.8.2-2 Table 6.8.2-5 [DELETED] ............................................................................................. 6.8.2-2 Table 6.8.2-6 Isotopic ketr Bias Uncertainty for Actinides and Fission Product Bumup Credit Evaluations of PWR SNF ............................................... 6.8.2-3 Table 6.10.1-1 PWR Nonfuel Insert Study (Guide and Instrument Tube Insertion) ... 6.10.1-83 Table 6.10.1-2 PWR Basket Component Tolerance and Shift Study Results (Independent Variations) ............................................................................................ 6.10.1-84 Table 6.10.1-3 PWR Basket Component Tolerance and Shift Study Results (Combined Variations; Radial In Shift) .................................................................. 6.10.1-85 Table 6.10.1-4 Depletion Analysis Analyzed Parameters ............................................ 6.10.1-86 Table 6.10.1-5 Axial Profile Shape for Depletion Analysis ......................................... 6.10.1-87 Table 6.10.1-6 Isotopes Extracted from TRITON Runs .............................................. 6.10.1-87 Table 6.10.1-7 TRITON Parameter Trends (Cask Analysis - NUREG/CR-6665) ..... 6.10.1-89 Table 6.10.1-8 NUREG "Reasonably" Bounding Parameters ..................................... 6.10.1-90 Table 6.10.1-9 Undamaged PWR Fuel - Fresh Fuel Maximum Enrichments ............. 6.10.1-91 Table 6.10.1-10 Undamaged PWR Fuel Hypothetical End-Drop Accident Evaluation ............................................................................................ 6.10.1-91 Table 6.10.1-11 [DELETED] ......................................................................................... 6.10.1-91 Table 6.10.1-12 Maximum Initial Enrichment- 37 Assembly Undamaged Fuel Configuration - 0.036 g/cm2 10B Absorber .......................................... 6.10.1-92 Table 6.10.1-13 Maximum Initial Enrichment- 37 Assembly Undamaged Fuel Configuration - 0.030 g/cm 2 1°B Absorber .......................................... 6.10.1-92 Table 6.10.1-14 Maximum Initial Enrichment-37 Assembly Undamaged Fuel Configuration - 0.027 g/cm 2 1°B Absorber .......................................... 6.10.1-93 Table 6.10.1-15 [DELETED] ......................................................................................... 6.10.1-93 NAC International 6-vi
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0 List of Tables (cont'd)
Table 6.10.1-16 Maximum Initial Enrichment - Undamaged Fuel Configuration WEIS - Optional Configurations ........................................................ 6.10.1-94 Table 6.10.1-17 RCCA Geometry .................................................................................. 6.10.1-94 Table 6.10.1-18 RCCA Calculated Isotopic Composition ............................................. 6.10.1-95 Table 6.10.1-19 RCCA Input Composition (atom/b-cm) ............................................... 6.10.1-96 Table 6.10.1-20 [DELETED] ......................................................................................... 6.10.1-96 Table 6.10.1-21 [DELETED] ......................................................................................... 6.10.1-96 Table 6.10.1-22 Undamaged Fuel Basket - RCCA Controlled / Spent Fuel Comparison .......................................................................................... 6.10.1-97 Table 6.10.1-23 [DELETED] ......................................................................................... 6.10.1-97 Table 6.10.1-24 [DELETED] ......................................................................................... 6.10.1-97 Table 6.10.1-25 [DELETED] ......................................................................................... 6.10.1-97 Table 6.10.1-26 [DELETED] ......................................................................................... 6.10.1-97 Table 6.10.1-27 [DELETED] ......................................................................................... 6.10.1-98 Table 6.10.1-28 [DELETED] ......................................................................................... 6.10.1-98 Table 6.10.1-29 [DELETED] ......................................................................................... 6.10.1-98 Table 6.10.1-30 Single Assembly Misload - Most Reactive Fuel Assemblies ............. 6.10.1-98 Table 6.10.1-30a Babcock & Wilcox Single Misload Assembly Results ........................ 6.10.1-99 Table 6.10.l-30b Combustion Engineering Single Misload Assembly Results .............. 6.10.1-99 Table 6.10.1-30c Westinghouse Single Misload Assembly Results .............................. 6.10.1-100 Table 6.10.l-30d 0.030 g/cm 2 10B Content - 75% Effective Absorber Sheet ............... 6.10.1-100 Table 6.10.1-30e 0.027 g/cm2 1°B Content- 67.5% Effective Absorber Sheet ............ 6.10.1-101 Table 6.10.1-31 Undamaged PWR Basket Reactivity Summary ................................. 6.10.1-102 Table 6.10.1-32 Power Density Bias Calculation for Depletion Analysis ................... 6.10.1-103 Table 6.10.1-33 Fission Product and Minor Actinide Worth and Bias ........................ 6.10.1-104 Table 6.10.1-34 Bumup Credit Depletion Bias Uncertainty ........................................ 6.10.1-105 Table 6.10.1-35 Bumup and Enrichment Ranges for Loading Curve Data ................. 6.10.1-106 Table 6.10.1-36 MCNP6.1 Validation -Range of Applicability for PWR Fresh and Spent Fuel Evaluations ................................................................ 6.10.1-107 Table 6.10.1-37 MCNP Criticality Results RCCA (Ag-In-Cd) ................................... 6.10.1-109 Table 6.10.1-38 USLSTA TS Method I Results for Correlated Variables at 5% MMS - RCCA (Ag-In-Cd) .......................................................... 6.10.1-109 Table 6.10.1-39 Reactivity Results for CE Absorber Material Effects in Fuel Depletion Analysis ............................................................................................... 6.10.1.109
- NAC International 6-vii
MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision O List of Tables (cont'd)
Table 6.10.1-40 Reactivity Results for CEAs Partially Inserted in Fuel Dpeletion Analysis .............................................................................................. 6.10.1-110 Table 6.10.1-41 Reactivity Results for CEAs Fully Inserted for 5 GWd/MTU in Fuel Depletion Analysis ............................................................................. 6.10.1-110 Table 6.10.1-42 Reactivity Results for Additional Insert Bands in Fuel Dpeletion Analysis .............................................................................................. 6.10.1-111 Table 6.10.1-43 Reactivity Results for Mixed-Mode Depletion in Fuel Depletion Analysis .............................................................................................. 6.10.1-112 Table 6.10.1-44 Reactivity Results for Additional Depletion Bands and Mixed-Mode Depletion in RCCA Depletion Analysis ............................................ 6.10.1-112 Table 6.10.2-1 Damaged to Undamaged Basket Reactivity Comparison .................... 6.10.2-10 Table 6.10.2-2 Damaged Fresh Fuel Mixture Height Reactivity Study -
0.036 g/cm2 1°B .................................................................................... 6.10.2-11 Table 6.10.2-3 Damaged Fresh Fuel Reactivity - 0.036 g/cm2 1°B ............................. 6.10.2-12 Table 6.10.2-4 Undamaged with Burnup Credit and Damaged Fuel Reactivity Comparison- 0.036 g/cm 2 10B ............................................................ 6.10.2-12 Table 6.10.2-5 Damaged Fuel Basket Loading Curve Summary - 0.036 g/cm2 1°B ... 6.10.2-13
- Table 6.10.2-6 Damaged Fuel Basket Loading Curve Summary - 0.030 g/cm2 1°B ... 6.10.2-13 Table 6.10.2-7 Damaged Fuel Basket Loading Curve Summary- 0.027 g/cm2 1°B ... 6.10.2-14 Table 6.10.2-8 Damaged Fuel Maximum Reactivity Moderation Study ..................... 6.10.2-15 Table 6.10.2-9 Damaged Fuel Missing Rod Study ...................................................... 6.10.2-16 Table 6.10.2-10 Undamaged Fuel Basket - RCCA Controlled/ Spent Fuel Comparison .......................................................................................... 6 .10 .2-16 Table 6.10.2-11 Maximum Initial Enrichment - Damaged Fuel Configuration WEl 5 - Optional Configurations ......................................................... 6.10.2-17 Table 6.10.2-12 Burnup and Enrichment Ranges for Damaged Basket Loading Curve Data ........................................................................................... 6.10.2-18 Table 6.10.3-1 Overpack Comparison Results - 87 Assemblies ................................. 6.10.3-59 Table 6.10.3-2 Overpack Comparison Results - 82 Assemblies ................................. 6.10.3-60 Table 6.10.3-3 System Reactivity Response to BWR Fuel Type and Pellet-to-Clad Condition .............................................................................................. 6.10.3-61 Table 6.10.3-4 System Reactivity Response to BWR Fuel Type and Channel Thickness ............................................................................................. 6.10.3-62 Table 6.10.3-5 BWR Lattice Parameter Reactivity Study (Increased Variance) ......... 6.10.3-63 Table 6.10.3-6 BWR Heterogeneous vs. Homogeneous Enrichment Analysis Results .................................................................................................. 6.10.3-64 NAC International 6-viii
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C List of Tables (cont'd)
Table 6.10.3-7 BWR 87-Assembly Basket Component Tolerance and Shift Study Results (Independent Variations) ......................................................... 6.10.3-66 Table 6.10.3-8 BWR 87-Assembly Basket Component Tolerance and Shift Study Results (Combined Variations; Radial In Shift) .................................. 6.10.3-67 Table 6.10.3-9 BWR 82-Assembly Basket Component Tolerance and Shift Study Results (Combined Variations; Radial In Shift) .................................. 6.10.3-68 Table 6.10.3-10 Tube Dimensioning/ Tolerancing Study Results ................................ 6.10.3-69 Table 6.10.3-11 Top End Drop Results, No Axial Blanket- 87 Assembly .................. 6.10.3-70 Table 6.10.3-12 Top End Drop Results, No Axial Blanket- 82 Assembly .................. 6.10.3-71 Table 6.10.3-13 Top End Drop Results, 6" Axial Blanket- 87 Assembly .................... 6.10.3-72 Table 6.10.3-14 Top End Drop Results, 6" Axial Blanket-82 Assembly .................... 6.10.3-73 Table 6.10.3-15 Top End Drop Results, 6" No Bale Crush -No Axial Blanket- 87 Assembly .............................................................................................. 6.10.3-74 Table 6.10.3-16 Top End Drop Results, 6" Axial Blanket - 82 Assembly .................... 6.10.3-75 Table 6.10.3-17 Top End Drop Results, No Blanket, Reduced Enrichment - 87 Assembly (0.027 g/cm2 1OB Absorber) .............................................. 6.10.3-76 Table 6.10.3-18 Top End Drop Results, No Blanket, Reduced Enrichment- 82 Assembly (0.027 g/cm2 1OB Absorber) .............................................. 6.10.3-77 Table 6.10.3-19 87 Assembly - Maximum Enrichment Fuel with and without Axial End Blanket (0.027 g/cm2 1OB Absorber) .......................................... 6.10.3-78 Table 6.10.3-20 82 Assembly - Maximum Enrichment Fuel with and without Axial End Blanket .......................................................................................... 6.10.3-79 Table 6.10.3-21 BWR Maximum Reactivity Moderation Results ................................. 6.10.3-80 Table 6.10.3-22 BWR Maximum System Reactivity Results (Infinite Array) .............. 6.10.3-80 Table 6.10.3-23 BWR Single Cask Result Summary .................................................... 6.10.3-81 Table 6.10.3-24 Maximum Allowed Enrichment - 0.0225 g/cm2 Effective 1OB -
87 Assembly ......................................................................................... 6.10.3-82 Table 6.10.3-25 Maximum Allowed Enrichment - 0.0225 g/cm2 Effective 1OB -
82 Assembly ......................................................................................... 6.10.3-83 Table 6.10.3-26 Maximum Allowed Enrichment - 0.020 g/cm2 Effective 1OB -
87 Assembly ......................................................................................... 6.10.3-84 Table 6.10.3-27 Maximum Allowed Enrichment - 0.020 g/cm2 Effective 1OB -
82 Assembly ......................................................................................... 6.10.3-85 Table 6.10.3-28 MCNP Validation - Range of Applicability for BWR Evaluations .... 6.10.3-86 Table 6.10 .4-1 PWR Moderator Exclusion Results ....................................................... 6.10.4-2 Table 6.10.4-2 BWR Moderator Exclusion Results ....................................................... 6.10.4-2
- NAC International 6-ix
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 6.1 Description of Criticality Design 6.1.1 Design Features MAGNATRAN consists of a TSC (Transportable Storage Canister) and a transport cask overpack. The system is designed to safely transport up to 37 PWR fuel assemblies, up to 87 BWR fuel assemblies, or GTCC materials. The system is also designed to transport up to four damaged fuel cans (DFCs) in the DF Basket Assembly. The DF Basket Assembly has a capacity ofup to 37 undamaged PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations. Each DFC may contain an undamaged PWR fuel assembly, a damaged PWR fuel assembly, or PWR fuel debris equivalent to one PWR fuel assembly. Undamaged PWR fuel assemblies may be placed directly in the DFC locations of a DF Basket Assembly.
The TSC is comprised of a stainless steel canister and a fuel basket. Both the PWR and BWR system include two TSC lengths to transport fuel assemblies. Spacers inside the TSC may be employed to facilitate loading or unloading operations. Spacer use is not required by the criticality analysis presented in this chapter. Axial movement evaluated does not credit the presence of spacers. Should spacers be used they stay in the TSC during transportation. Fuel is
- loaded into the TSC contained within a transfer cask under water in the spent fuel pool. Once loaded with fuel, the TSC is drained, dried, backfilled with helium and welded closed. The welded TSC boundary is designed to withstand all normal conditions and hypothetical accident events and to retain a no credible leakage boundary. A single transport cask accommodates all of the PWR and BWR TSCs. An axial spacer is used inside the transport cask cavity for the short TSCs.
Based on a no credible leakage TSC boundary and a leaktight transport cask boundary, moderator is not present in the TSC while it is being transported. The structural evaluations of the MAGNATRAN cask demonstrate that the cask containment is maintained during all conditions of transport. Containment boundary integrity is checked via leakage tests described in Section 8.2.2. With no credible leakage into the fissile material region, the under moderated system will remain significantly subcritical during all normal and hypothetical accident conditions.
Evaluations in this chapter address allowable payload limitations when invoking moderator exclusion and for configurations when the cask containment an TSC interiors are flooded in the criticality evaluations. Discussion in Sections 6.2 to 6.6 are relevant to the flooded TSC configuration. Moderator exclusion evaluations are included in Section 6.10. Maximum reported reactivity for moderator exclusion is simply a function of maximum fissile material
- mass and initial enrichment.
NAC International 6.1.1-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C System criticality control is achieved through a combination of neutron absorber sheets on the interior faces of the fuel tubes and in the case of moderator intrusion for the PWR system the use of actinide-and fission product bumup credit.
Individual fuel assemblies are supported in place by the fuel tubes, by developed cells formed by the fuel tubes, or by a combination of fuel tubes and side or comer weldments. The baseline neutron absorber modeled is a borated aluminum sheet with effective 1°13 loadings of 0.036 g/cm2 and 0.027 g/cm 2 for the PWR and BWR system, respectively. The system is also evaluated for effective 1°13 loadings of 0.030 and 0.027 g/cm 2 for PWR baskets and 0.0225 and 0.020 g/cm2 for BWR baskets. The minimum as-manufactured loading of the neutron absorber sheets depends on the effectiveness of the absorber and the minimum effective absorber areal density. Effectiveness of the absorber is influenced by the uniformity and quantity of the 1°B nuclide within the absorber base material. Depending on the absorber type, a 75% or 90%
effectiveness is credited. Any material meeting the 1°B areal density and physical dimension requirements will produce similar reactivity results. See Table 6.1.1-1 for effective versus "credit" adjusted absorber areal densities.
A combination of stainless steel cover sheets and weld posts holds the neutron absorber sheets in place. The PWR basket design includes 21 fuel tubes forming 37 fuel-assembly-sized openings, while the BWR basket contains 45 fuel tubes forming 89 fuel-assembly-sized openings. The PWR damaged fuel basket design includes 17 fuel tubes and four comer weldments forming 37 openings. A sketch of a cross-section of the damaged fuel basket is shown in Figure 6.1.1-2.
The combination of 45 BWR fuel tubes with four comer and four side weldments forms 89 fuel-assembly-sized openings; however, two openings are below the vent and drain ports and are not loaded. For simplicity and cask symmetry, all 89 slots are modeled as filled with fuel. An optional "82-assembly" configuration of the BWR basket is evaluated, where five center openings in an X" pattern are left unoccupied (the basket model fills the openings below the port cover and, therefore, contains 84 assemblies). See Figure 6.1.1-1 for the loadable basket locations in the 82-assembly basket configuration.
NAC International 6.1.1-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 6.1.2 Summary Criticality Evaluation Results System reactivity was evaluated for maximum payloads of 37 PWR assemblies in an undamaged fuel basket, 89 BWR assemblies, 37 PWR assemblies in a damaged fuel basket (includes four DFCs), and GTCC materials.
Key assembly physical characteristics for each PWR and BWR fuel assembly type are shown in Table 6.1.2-1 for the PWR systems and Table 6.1.2-2 for the BWR system. System reactivity is evaluated for each of the listed fuel assembly types.
For moderated systems BWR maximum enrichments for the 82-assembly basket and the 87-assembly basket are listed in Table 6.1.2-3 for fuel with axial unenriched end-blankets and in Table 6.1.2-4 for fuel without unenriched axial end blankets. Burnup credit loading curves for the PWR system considering moderator are included in Section 6.10.1 for the undamaged fuel basket and the damaged fuel basket not containing damaged fuel and in Section 6.10.2 for the damaged fuel basket containing damaged fuel.
Considering moderator in the TSC the PWR WEIS under-burned fuel assemblies may be transported at a maximum fuel enrichment of 4.05 wt% 235 U and minimum burnup of 12,000 MWd/MTU provided a Ag-In-Cd full-length RCCA is inserted into the fuel assembly. The RCCA used for reactivity suppression may be new or irradiated up to an equivalent exposure of 200,000 MWd/MTU. Loading of the under-burned assemblies, each with their own RCCA, is limited to the center nine fuel storage locations.
Under conditions of moderator exclusion from the fissile material region, i.e., the TSC interior, a 5 wt°/o 235 U initial enrichment is allowed to be loaded for all PWR and BWR fuel types. No burnup credit is required under this condition.
PWR results represent the bounding values for fuel assemblies with and without nonfuel inserts in the guide tubes. Maximum enrichment is defined as peak rod enrichment (maximum pellet enrichment) for PWR assemblies and the maximum peak planar-average enrichment for BWR assemblies. The maximum initial peak planar-average enrichment is the maximum planar-average enrichment at any height along the axis of the fuel assembly.
Undamaged assemblies are evaluated with a full, nominal set of fuel rods. Fuel rod (lattice) locations may contain filler rods. A filler rod must occupy, at a minimum, a volume equivalent to the fuel rod it displaces. Filler rods may be placed into the lattice after assembly in-core use or be designed to replace fuel rods prior to use, such as integral burnable absorber rods .
- NAC International 6.1.2-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C An undamaged assembly must contain its nominal set of guide and instrument tubes (PWR), and water rods (BWR). Analysis demonstrated that variations in the guide/instrument tube and water rod thickness and diameter have no significant effect on system reactivity.
Maximum reactivities for package and arrays of PWR undamaged fuel, PWR damaged fuel, and BWR undamaged systems are discussed in Section 6.10.1, 6.10.2 and 6.10.3 respectively.
Lower maximum reactivities were calculated for moderator exclusion systems at maximum enrichments. BWR evaluations rely on a single upper safety limit (USL) and have the maximum reactivity (keff) case reported. PWR analysis rely on various USLs as depletion code bias/bias uncertainty is affect by assembly burnup. PWR burnup and initial enrichment load limitations are calculated to result in maximum reactivity below the applicable USLs. There is not a single maximum system reactivity (keff) point to report. The analysis demonstrates that all allowed payload configurations are below reactivity limits. GTCC material does not have significant fissile material content and is subcritical under all conditions.
NAC International 6.1.2-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 6.1.2-1 PWR Fuel Assembly Loading Criteria (Assembly Description)
Min Min Max Max No. of No. of Max Clad Clad Pellet Active Max Assembly Fuel Guide Pitch OD Thick. OD Length Load Type Rods Tubes* (Inch) (Inch) (Inch) (Inch) (inch) (MTU)
BW15H1 208 17 0.568 0.43 0.0265 0.3686 144.0 0.4858 BW15H2 208 17 0.568 0.43 0.025 0.3735 144.0 04988 BW15H3 208 17 0.568 0.428 0.023 0.3742 144.0 0.5006 BW15H4 208 17 0.568 0.414 0.022 0.3622 144.0 0.4690 BW17H1 264 25 0.502 0.377 0.022 0.3252 144.0 0.4799 CE14H1 176 5 0.58 0.44 0.026 0.3805 137.0 0.4167 CE16H1 236 5 0.5063 0.382 0.025 0.325 150.0 0.4463 WE14H1 179 17 0.556 0.40 0.0162 0.3674 145.2 0.4188 WE15H1 204 21 0.563 0 422 0.0242 0 3669 144.0 0.4720 WE15H2 204 21 0.563 0.417 0.0265 0 357 144.0 0.4469 WE17H1 264 25 0 496 0.372 0.0205 0.3232 144.0 0.4740 WE17H2 264 25 0.496 0.36 0.0225 0.3088 144.0 0.4327 Assembly characteristics represent cold, unirradiated, nominal configurations.
Fuel assembly load curves are provided per general core/lattice size (e.g., all four BW15Hx fuel types are combined under the BWl 5 curve).
a Combined number of guide and instrument tubes .
- NAC International 6.1.2-3
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 6.1.2-2 BWR Fuel Assembly Loading Criteria (Assembly Description)
Min Min Max Max Number Number of Max Clad Clad Pellet Active Max Assembly of Fuel Partial Pitch OD Thick. OD Length Loading Type Rods Len!:lth Rods (inch) (inch) (inch) (inch) (inch) (MTU)
B7 48A 48 N/A 0.7380 0.5700 0.03600 0.4900 144.0 0.1981 B7 49A 49 N/A 0.7380 0.5630 0.03200 0.4880 146.0 0.2034 B7 49B 49 N/A 0.7380 0.5630 0.03200 0.4910 150.0 0.2115 B8 59A 59 N/A 0.6400 0.4930 0.03400 0.4160 150.0 0.1828 B8 60A 60 N/A 0.6417 0.4840 0.03150 0.4110 150.0 0.1815 B8_60B 60 N/A 0.6400 0.4830 0.03000 0.4140 150.0 0.1841 B8 61B 61 N/A 0.6400 0.4830 0.03000 0.4140 150.0 0.1872 B8 62A 62 N/A 0.6417 0.4830 0.02900 0.4160 150.0 0.1921 B8 63A 63 N/A 0.6420 0.4840 0.02725 0.4195 150.0 0.1985 B8 64A 64 N/A 0.6420 0.4840 0.02725 0.4195 150.0 0.2017 B8 64B 64 N/A 0.6090 0.4576 0.02900 0.3913 150.0 0.1755 B9 72A B9_74A B9 76A B9 79A B9 BOA 72 74 8 76 79 N/A 8
N/A N/A N/A 0.5720 0.5720 0.5720 0.5720 0.4330 0.4240 0.4170 0.4240 0.02600 0.02390 0.02090 0.02390 0.3740 0.3760 0.3750 0.3760 150.0 150.0 150.0 150.0 0.1803 0.1873 0.1914 0.2000 80 0.5720 0.4230 0.02950 0.3565 150.0 0.1821 B10 91A 91 8 8 0.5100 0.3957 0.02385 0.3420 150.0 0.1906 B10 92A 92 8 14 0.5100 0.4040 0.02600 0.3455 150.0 0.1966 B10 96A 96 8 12 0.4880 0.3780 0.02430 0.3224 150.0 0.1787 B10_100A 100 N/A 0.4880 0.3780 0.02430 0.3224 150.0 0.1861 Note: Assembly characteristics represent cold, unirradiated, nominal configurations.
a Assemblies may contain partial-length fuel rods. Partial-length rod assemblies are evaluated by removing partial-length rods from the lattice. This configuration bounds an assembly with full-length rods and combinations of full- and partial-length rods.
NAC International 6.1.2-4
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 6.1.2-3 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel With Axial Blanket)
Max. Initial Enrichments ( wt % 235U)
Absorber" 0.027 g/cm2 10s Absorber" 0.0225 g/cm21os Absorber" 0.02 gtcm21os Fuel 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Type Basket Basket Basket Basket Basket Basket 87 48A 4.0% 4.5% 3.7% 4.5% 3.6% 4.4%
87 49A 3.8% 4.5% 3.6% 4.4% 3.5% 4.3%
87_498 3.8% 4.5% 3.6% 4.4% 3.5% 4.2%
88 59A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%
88 60A 3.8% 4.5% 3.7% 4.4% 3.5% 4.2%
88 608 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
88 618 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
88 62A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%
88 63A 3.8% 4.5% 3.6% 4.3% 3.4% 4.2%
88 64A 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%
88 648 3.6% 4.3% 3.4% 4.1% 3.3% 4.0%
- 89 72A 89 74A 89 76A 89 79A 89 80A 3.8%
3.7%
3.5%
3.7%
3.8%
4.5%
4.3%
4.2%
4.4%
4.5%
3.6%
3.4%
3.4%
3.4%
3.6%
4.3%
4.1%
4.0%
4.2%
4.3%
3.4%
3.4%
3.3%
3.3%
3.5%
4.1%
4.0%
3.9%
4.0%
4.2%
810 91A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
810 92A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%
810 96A 3.7% 4.3% 3.5% 4.1% 3.4% 4.0%
810_100A 3.6% 4.4% 3.5% 4.1% 3.4% 4.0%
a Maximum planar average.
b Borated aluminum neutron absorber sheet effective areal 1
°13 density .
- NAG lnternatlonal 6.1.2-5
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 6.1.2-4 Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel Without Axial Blanket)
Max. Initial Enrichmenta (wt% 235 U)
Absorberb 0.027 i:i/cm2 108 Absorberb 0.0225 Afcm 210 8 Absorberb 0.02 Q/cm2 108 Fuel 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Type Basket Basket Basket Basket Basket Basket 87 48A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%
87 49A 3.7% 4.5% 3.6% 4.3% 3.4% 4.1%
87 498 3.7% 4.5% 3.6% 4.3% 3.5% 4.2%
88 59A 3.8% 4.5% 3.7% 4.4% 3.5% 4.3%
88 60A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
88 608 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
88 618 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%
88 62A 3.6% 4.4% 3.5% 4.2% 3.4% 4.1%
88 63A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
88 64A 3.7% 4.5% 3.5% 4.3% 3.4% 4.1%
88 648 3.6% 4.2% 3.4% 4.1% 3.3% 4.0%
89 72A 89 74A 89 76A 89 79A 89 BOA 3.7%
3.6%
3.5%
3.5%
3.7%
4.4%
4.2%
4.1%
4.2%
4.5%
3.5%
3.4%
3.3%
3.4%
3.6%
4.2%
4.1%
4.0%
4.1%
4.3%
3.4%
3.3%
3.2%
3.2%
3.5%
4.1%
4.0%
3.8%
3.9%
4.1%
810 91A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%
810 92A 3.7% 4.4% 3.6% 4.2% 3.4% 4.1%
810_96A 3.6% 4.2% 3.4% 4.1% 3.4% 4.0%
810_100A 3.6% 4.3% 3.4% 4.0% 3.3% 3.9%
Table 6.1.2-5 [DELETED]
[Deleted]
a Maximum planar average.
b Borated aluminum neutron absorber sheet effective areal 10B density.
NAC International 6.1.2-6
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 6.4.1 Configuration/Discussion Transport regulations for fissile material are codified in IO CFR 71. The following paragraphs demonstrate that the configuration evaluated meets each of the licensing requirements by listing the requirement and then providing a MAGNATRAN-specific response.
"71.55 (b) &cept as provided in paragraph (c) or (g) of this section, a package used for the shipment offissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained."
The MAGNATRAN system is evaluated in two configurations. The first includes moderator intrusion into the fissile material region. The second applies moderator exclusion and requires the application of 10 CFR 71.55 (c).
"71.55 (c) The Commission may approve exceptions to the requirements ofparagraph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to
- ensure that the containment system does not leak."
The MAGNATRAN system is designed with two (2) independent boundaries to prevent moderator intrusion into the fissile material region, the transport cask and the TSC. The TSC is weld sealed and, as described in Section 2.1.1.3, analyses show that the TSC maintains its leaktight confinement function in all of the evaluated conditions. The structural evaluations of the MAGNATRAN cask demonstrate that the cask containment is maintained during all conditions of transport. Leakage testing of the transport cask is documented in Chapter 8. The confinement boundary of the TSC is discussed within the IO CFR 72 license.
Under moderator intrusion condition the MAGNATRAN maximum reactivity configuration PWR and BWR package configuration is evaluated by removing the reflecting boundary condition from the maximum reactivity array configuration, retaining a maximum reactivity physical system configuration at various moderator conditions (full density water and void in the TSC interior). Moderator density curves provided in the array study documents that intermediate moderator densities are bounded by the full density case.
"7I.55(b)(2)(1) The most reactive credible configuration consistent with the chemical and physical form of the material. "
- NAC International 6.4.1-1
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C The chemical and physical form of the fissile material is solid uranium oxide powder in zirconium alloy rods. All credible configurations of the fuel rods are evaluated. Structural analysis ensures that cask operating conditions do not result in rod configurations outside the evaluated conditions.
During moderator exclusion evaluations the system was evaluated with the basket removed and a homogeneous sphere of UOi filling the inner diameter of the canister cavity. The sphere represents a maximum reactivity, minimum neutron leakage, configuration.
Damaged fuel under TSC flooded conditions is addressed in various configurations, including dispersed fuel pellets in water. As the fuel pellets are significantly higher in density (~ 10.5 g/cm 3) than moderator(~ 1 g/cm 3), this configuration bounds that of any credible physical configuration of the fissile material within moderator.
"(2) Moderation by water to the most reactive credible extent."
As discussed in previous text, per 71.55 (c), the design of the MAGNATRAN system precludes moderator from entering the system. Evaluations are included for configuration at moderator conditions representing maximum reactivity and for moderator extrusion from the TSC.
"(3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging. " A single package reflected at either the containment boundary, or at the intact cask boundary, by 20 cm of water demonstrates that system reactivity does not increase above those calculated for the cask arrays.
"71.55(d) A package used for the shipment offissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71. 71 ("Normal conditions of transport)
"(1) The contents would be subcritical;"
MAGNATRAN maximum reactivity is below the USL under all normal conditions of transport as demonstrated by both single cask and cask array analyses.
"(2) The geometric form of the package contents would not be substantially altered;"
Structural analysis confirms that no normal condition of transport will result in a substantially altered package.
"(3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under§ 71.59(a)(l), it has been assumed that moderation is present to NAC International 6.4.1-2
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and" Criticality evaluations are provided in two configurations. The first applies water in-leakage into containment. The second applies a no credible leakage TSC boundary and the leaktight cask boundary to assure no moderator intrusion into the TSC.
"(4) There will be no substantial reduction in the effectiveness of the packaging, including:
"(i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed;
"(ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer sw-face of the packaging; and
"(iii) No occurrence of an aperture in the outer surface of the packaging large enough to permit the entry of a 10 cm (4 in) cube. "
Structural analysis confirms that no normal condition of transport will result in a reduced effectiveness of the as-evaluated package (analysis accounts for a potential decrease in the neutron shield after the hypothetical accident condition fire). There is no significant reduction in package volume and spacing of fissile material, and there is no aperture on the outer surface of the package under any normal condition of transport.
"(e) A package used for the shipment of.fissile material must be so designed and constructed and its contents so limited that under the tests specified in§ 71. 73 ("Hypothetical accident conditions), the package would be subcritical. For this determination, it must be assumed that:
"(]) The fissile material is in the most reactive credible configuration consistent with the damaged condition of the package and the chemical and physical form of the contents; "
Impact limiters are installed on the cask to ensure that the package retains its pre-accident configuration through all accident condition drops. The limiters restrict acceleration loads on the package and contents. The cask neutron shield is removed for accident condition criticality analysis. Fire accident temperatures exceed the material allowable. Therefore, the analysis conservatively removed the neutron shield from the analysis models.
"(2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and" The post-accident configuration cask retains its ability to exclude water from the containment boundary. Furthermore, as demonstrated by structural analysis, the welded TSC boundary
- remains in a no credible leakage condition through all accident conditions.
NAC International 6.4.1-3
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
"(3) There is full reflection by water on all sides, as close as is consistent with the damaged condition of the package. "
The damaged cask configuration retains the outer cask shell. This cask configuration is the baseline for all criticality analysis in this chapter.
NAC International 6.4.1-4
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- NAC International 6.10.4-1
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Table 6.10.4-1 PWR Moderator Exclusion Results Table 6.10.4-2 BWR Moderator Exclusion Results NAC International 6.10.4-2
"NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Figure 6.10.4-1 Homogeneous U02 Sphere - VISED XZ Slice
- NAC International 6.10.4-3
MAGNATRAN Transport Cask SAR Aprll 2019 Docket No. 71-9356 Revision 0 7.1.2 Loading of Contents Note: The MAGNA1RAN transport cask is dry loaded with a fuel and GTCC waste TSC directly following TSC loading and closure, or following a period of on-site storage in the spent fuel building or facility, or at the onsite ISFSI using the MTC and attendant support hardware.
Operation of the MTC is described in the approved site-specific procedures and the MAGNASTOR@ FSAR. Site-specific procedures shall comply with the requirements of the SAR. Potential alternate procedures and site-specific hardware are described when necessary.
- 1. Install appropriate work platforms, scaffolding or lifts to allow access to the top of the MAGNA1RAN transport cask.
- 2. Detorque and remove the lid port coverplate bolts and store the coverplate and associated bolts to prevent damage.
- 3. Detorque the cask lid bolts in the reverse order of the torquing sequence indicated on the lid.
- 4. Remove the lid bolts, inspect the bolts for thread damage and store to prevent damage.
- 5. Install the two lid alignment pins in their designated threaded holes (#s 14 and 36) and hand-tighten.
6 . Install and tighten the swivel hoist rings (or equivalent approved site-specific lifting system)
- in the four threaded lifting holes in the cask lid and torque to the values specified in Table 7.1-1.
- 7. Attach an appropriate lifting sling set to the swivel hoist rings and to a crane hook. Lift the lid from the cask and store it to prevent damage.
- 8. Remove the two alignment pins. Using a crane and suitable slings, install the transfer shield ring into the lid recess.
Note: The transfer shield ring aligns the transfer cask adapter to the cask cavity, provides additional side shielding and protects the cask lid seating surface from damage.
Note: The following loading procedures are based on the dry transfer of a loaded and closed TSC containing spent fuel assemblies or a loaded and closed TSC containing GTCC waste either immediately following loading or following a period of interim storage.
All TSCs shall be independently verified to be in compliance with the CoC content conditions.
Note: An evaluation of TSCs containing spent nuclear fuel shall be performed to verify that the installed neutron absorbing materials required to assure criticality safety are acceptable for transport conditions.
Note: TSCs containing spent nuclear fuels that are to be retrieved from storage for off-site transport in the MAGNA TRAN transport cask will be evaluated to ensure that the specific TSC stored in the storage overpack, which may have been subject to 10 CFR NAC International 7.1-5
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 72 normal and off-normal, accident and natural phenomena events, retains its ability to satisfy functional and performance requirements of the MAGNATRAN packaging certified content conditions. Dry storage systems that have been maintained within an Aging Management Program will include system specific review and assessment of this information record as part of the off-site transport evaluation to ensure that the MAGNATRAN packaging certified content conditions are validated.
TSCs containing spent nuclear fuel and experiencing only normal or off-normal events during storage will be evaluated for potential corrosion at the welds and any damage caused by removal of the TSC from the storage overpack.
In addition to the evaluation done for normal/off-normal storage, TSCs containing spent nuclear fuel that have experienced accident or natural phenomena events must be evaluated for potential degradation of the fuel, basket, and neutron absorbers. This evaluation will be performed for each TSC as part of the preparation for loading for off-site transport using: 1) the annual inspection and surveillance records and off-normal and accident event reports that are maintained by the licensee for each loaded MAGNASTOR system in compliance with 10 CFR 72 requirements; and 2) in the case of storage accidents and natural phenomena events, any necessary examinations performed at the time of transfer to ensure the condition of the TSC and contents.
- TSC loading into the MAGNATRAN transport cask will be observed by operations staff noting any system interferences that occur during TSC retrieval from the storage overpack and during placement of the TSC into the transport cask. The cause of the interference and potential damage caused by the interference will be determined prior to shipment. Noted interferences will be made part of the TSC evaluation record to the extent required to validate that MAGNATRAN packaging content conditions are satisfied when the spent fuel canister is placed within the MAGNATRAN transport cask containment boundary for off-site transport.
- 9. The following procedures apply to fuel and GTCC waste TSC loading into the MAGNA TRAN transport cask after an on-site storage period or immediately following TSC loading:
9.a. For TSCs to be loaded in the MAGNATRAN transport cask following storage operations, remove the loaded TSC from the concrete cask (CC) and close the MTC shield doors. Record the time the TSC is lifted off the CC pedestal. Install the shield door locking devices.
9.b. For TSCs to be loaded immediately following loading and closing, prepare the TSC for transfer operations. Record the time the ACWS cooling of fuel TSC is terminated .
NAC International 7.1-6
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C Caution: In order to ensure that the spent fuel clad temperatures do not exceed 400°C in accordance with ISG-11, Revision 3, the time allowed for transfer of a loaded TSC containing spent fuel to the MAGNA TRAN cask is limited. The following time limits apply as noted:
Condition 1) For the maximum transportable fuel TSC heat loads of 23 kW for PWR and 22 kW for BWR, the maximum time from lifting the TSC off the CC pedestal (Section 7 .1.2, Step 9) for placement in the MTC through completion of the preparation of the MAGNA TRAN for transport and placement in a horizontal orientation on the transport vehicle (Section 7.1.3, Step 4) shall be< 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br />; Condition 2) For maximum heat load fuel TSCs loaded and closed immediately prior to loading into the MAGNATRAN cask, the maximum time from completion ofTSC closure operations, including helium backfill time and termination of external cooling (for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of the TSC (Section 7.1.4, Step 17) through completion of the preparation of the MAGNA TRAN for transport and placement in a horizontal orientation on the transport vehicle (Section 7.1.3, Step 4), shall be< 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />.
Note: These maximum transfer and preparation times are not applicable to the loading of GTCC waste TSCs as the ISG-11 temperature limits are not applicable.
- Note: In the event that the transfer and MAGNATRAN preparation procedures through placement of the cask in a horizontal orientation are not completed within the specified time period, corrective actions shall be implemented to return the TSC to the MAGNASTOR Transfer Cask (MTC) where active cooling of the TSC can be completed in accordance with procedures established in the MAGNASTOR FSAR and Operating Manual. The corrective actions shall be implemented with sufficient time to ensure that the maximum transfer and preparation times are not exceeded.
The external cooling of the TSC shall be continued for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to reduce the fuel clad and TSC internal component temperatures to allow re-start of MAGNATRAN loading procedures. The maximum time limit is 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> for both Condition 1 and 2 for each subsequent TSC transfer and transport preparation activity.
Note: The time limits specified in Step 9 are for the maximum allowable heat loads in the MAGNATRAN transport cask. Although transfer and cask preparation times would be longer for lower content decay heat loads, the limits for the maximum heat loads (PWR - 23 kW; BWR - 22 kW) will be conservatively implemented to all content decay heat loads .
- NAC International 7.1-7
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- 10. Connect lift slings to the transfer adapter, lift the adapter, and place it on the top of the MAGNATRAN cask. Visually verify proper fit-up with the transfer shield ring positioned in the lid recess.
- 11. Connect and verify operation of the transfer adapter auxiliary hydraulic system.
- 12. Install the TSC lifting hoist rings and sling set (or equivalent TSC lifting system meeting the facility's heavy load program) in the TSC closure lid threaded holes. Torque the hoist rings to the torque specified in Table 7.1-1.
- 13. Using the MTC lift yoke, engage the lifting trunnions and position the MTC containing the loaded TSC on the transfer adapter positioned on the MAGNATRAN transport cask.
Remove the shield door stops.
- 14. Install a stabilization system for the MTC, if required by the facility heavy load handling or seismic analysis programs.
- 16. Connect TSC sling set(s) (or site-specific approved TSC lifting system meeting the facility's heavy load program) to the crane hook. Verify that the MTC retaining components are in the engaged position.
- 17. Lift the TSC off the MTC shield doors (approximately1/2 inch) and open the doors using the auxiliary hydraulic system.
18, Lower the loaded TSC into the MAGNATRAN cask until the TSC rests on the bottom ofthe cask cavity.
- 19. Disengage the lifting sling set(s) from the hook or disengage the site-specific approved lifting system meeting the facility's heavy load program from the TSC. Close the MTC shield doors and install the door stops.
- 20. Retrieve the MTC lift yoke, engage the lifting trunnions, remove the stabilization system (if used), and lift the MTC off the top of the MAGNATRAN cask. Move the MTC and lift yoke from the area and store.
- 21. Disconnect the auxiliary hydraulic system connections, attach lifting slings to the transfer adapter and lift and move the transfer adapter from the area and store.
- 22. Remove the TSC lifting sling set(s), hoist rings or other site-specific approved TSC lifting system components from the top of the TSC.
- 23. Attach lifting slings, lift and remove the transfer shield ring. Inspect the cask lid O-ring seating surface for cleanliness and integrity.
- 24. Engage the cask lid lifting sling set to the crane and position the lid for seal inspection and replacement.
NAC International 7.1-8
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- 25. Remove and replace the lid metallic O-ring with an approved spare. Inspect the lid outer O-ring seal and, if it is damaged, replace it with an approved spare. Ensure that the replaced O-rings are properly installed and seated.
- 26. For the transport of shorter length TSCs install the cask cavity spacer to the bottom side of the lid with the four attachment bolts and lock washers. Torque the attachment bolts to the value specified in Table 7.1-1.
- 27. Install the two lid alignment pins in their designated threaded hole locations (#s 14 and 36) in the cask and hand-tighten.
- 28. Lift the cask lid and position the lid to engage the alignment pins. Slowly lower the lid into position. Remove the lid lifting equipment and alignment pins.
- 29. Inspect the 48 lid bolts for damage and replace, as required, with approved spares. Lubricate the bolts with nuclear-grade Never-Seeze, or equivalent, and install the lid bolts to hand tight.
In a minimum of four passes of increasing torque, torque the 48 lid bolts to the final value specified in Table 7.1-1 following the torquing sequence pattern marked on the lid.
- 30. Connect a vacuum pump and helium gas backfill system to the lid port quick-disconnect valve. Evacuate the cask cavity to a vacuum pressure of< 3 torr and backfill the cavity with helium gas to a pressure of 17.5, +2.5,-0 psia .
- 31. Disconnect the vacuum pump and backfill system from the lid port quick-disconnect valve.
- 32. Remove and replace the port coverplate metallic O-ring with an approved spare. Inspect the coverplate outer O-ring seal and, if damaged, replace it with an approved spare. Ensure that the replaced O-rings are properly installed and seated.
- 33. Install the coverplate in the lid port recess, purge the volume under the coverplate with helium prior to final seating, and torque the coverplate bolts to the value specified in Table 7.1-1.
- 34. Remove the cask lid test port plug and connect a Helium Mass Spectrometer Leak Detector (He MSLD) to the test port to perform the helium leakage test.
- 35. Using the He MSLD evacuate the volume between the cask lid metallic inner O-ring and the outer O-ring to a vacuum of< 0.1 torr. Operate the He MSLD to detect for helium in the evacuated volume to confirm the lid leakage rate is less than or equal to 2.0 x 10-7 cm3/s, helium with a minimum test sensitivity of 1.0 x 10-7 cm 3 Is, helium.
- 36. Upon completion of an acceptable lid helium leakage test, disconnect the He MSLD from the lid test port connection. Replace test port plug metallic seal with an approved spare, install the lid test port plug and torque to the value specified in Table 7 .1-1 .
- NAC International 7.1-9
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- 37. Remove the coverplate test port plug and connect the He MSLD to the port to perform the helium leakage test.
- 38. Using the He MSLD evacuate the volume between the coverplate metallic inner O-ring and the outer O-ring to a vacuum of< 0.1 torr. Operate the He MSLD to detect for helium in the evacuated volume to confirm the lid port coverplate leakage rate is less than or equal to 2.0 x 10*7 cm3/s, helium with a minimum test sensitivity of 1.0 x 10-7 cm 3/s.
Note: If a helium leakage rate exceeding the specified leakage rate is measured during either the lid or lid port coverplate leakage test, determine the source of the leak, repair (i.e., replace O-ring with approved spare, clean seating surfaces), reinstall and torque, and retest to original acceptance criteria.
- 39. Upon completion of an acceptable lid port coverplate leakage test, disconnect the He MSLD from the coverplate test port connection. Replace test port plug metallic seal with an approved spare, install the coverplate test port plug and torque to the value specified in Table 7.1-1.
7.1.3 Preparation for Transport
- 1. Remove scaffolding or work platforms from around the top of the MAGNA TRAN transport cask.
- 2. Connect the cask lift yoke to the cask handling crane and engage the lift yoke arms to the two transport cask lifting trunnions.
- 3. Lift the cask and move it into position over the transport frame/vehicle rear support structure.
- 4. Lower the cask to engage the rotation trunnions in the transport frame/vehicle rear support structure, rotate the cask to the horizontal orientation and disengage the lift yoke.
Note: The rotation trunnions are offset from the centerline to ensure rotation of the cask in the proper direction.
- 5. Using slings and a suitable crane, detorque and remove the nine lifting trunnion attachment bolts from one trunnion. Remove, clean and inspect the trunnion and bolts, observing for damage/wear. Repeat for the second lifting trunnion and bolts and store the trunnions and trunnion bolts.
- 6. Clean and inspect the trunnion recess and bolt holes. Position and install the trunnion plug.
Apply nuclear-grade lubricant, such as Never-Seeze, or equivalent, to the three trunnion plug bolts, and torque the bolts to the value specified in Table 7 .1-1. Repeat for the second trunnion recess and plug.
- 7. Decontaminate and clean the surfaces of the transport cask to ensure removable contamination limits for transport are met.
NAC I nternatlonal 7.1-10
MAGNATRAN Transport Cask SAR December 2020
- Docket No. 71-9356
- 8. Install the transport tie-downs over the top forging and on the rotation trunnions.
Revision 20C
- 9. Using the impact limiter lifting sling set, lift and install the lower (rear) impact limiter.
Install retaining rods and attachment and jam nuts, and torque all the components to the values specified in Table 7.1-1. Install the retaining rod lock wires.
- 10. Repeat the installation sequence for the upper (front) impact limiter.
- 11. Install a tamper indicating device (TID) to the upper impact limiter attachment rods. Record the TID identification numbers on the cask loading checklist and shipping papers.
- 12. Perform final visual inspection of the cask to ensure proper package assembly in accordance with the CoC.
- 13. Complete radiation and contamination surveys of the package external surfaces and record the data. Ensure removable contamination and radiation dose rates survey results comply with the limits specified in 10 CFR 71.87(i) and G) to verify removable contamination levels are in compliance with 49 CFR 173 .443 and radiation dose rates comply with 10 CFR 71.4 7, respectively.
- 14. Measure the dose rate in millirems per hour at one meter from the package surface to determine the Transport Index (TI). Indicate the TI on the Radioactive Material labels
- applied to the package in accordance with 49 CFR 172, Subpart E .
- 15. Determine the appropriate Criticality Safety Index (CSI) assigned to the package contents in accordance with the CoC, and indicate the correct CSI on the Fissile Material label applied to the package per 49 CFR 172, Subpart E.
- 16. Measure and record the temperature of the MAGNATRAN shield shell outer surface at the top center of the package in the horizontal position and verify that the measured package external temperature meets the exclusive use shipment temperature limit of 85°C (185°F) per 10 CFR 71.43(g).
- 17. Install the personnel barrier to the transport vehicle and verify all access points are locked.
- 18. Perform a radiation survey of the package and the transport vehicle to assure radiation levels are in compliance with 49 CFR 173 .441.
- 19. Apply placards to the transport vehicle in accordance with 49 CFR 172 Subpart F.
- 20. Review the cask loading inspection checklist and verify all required steps, tests and verifications have been satisfactorily completed, and all routine determinations have been satisfactorily completed in accordance with 10 CFR 71.87.
- 21. Complete the shipping documentation for an Exclusive Use Shipment and provide special instructions to the carrier to maintain an Exclusive Use Shipment.
- 22. The MAGNATRAN transport cask (package) is now ready for release and transport .
NAC International 7 .1-11
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 7.1.4 Loading the Transportable Storage Canister with Spent Fuel Contents
- 1. Visually inspect the TSC to ensure that it is clean and free of debris.
- 2. Place the TSC in the MTC.
- 3. Place the MTC containing the TSC into the spent fuel pool.
- 4. Load the previously designated fuel assemblies into the TSC in accordance with applicable CoC requirements.
Note: The fuel assemblies and authorized nonfuel hardware shall be selected in compliance with the requirements of the approved contents specified in the CoC including limitations on fuel assembly, damaged fuel assemblies in Damaged Fuel Cans (DFCs), and nonfuel hardware positions within the basket. Assembly, DFC, and nonfuel hardware selection and placement within the basket shall be independently verified.
Note: Verification of the location of high reactivity fuel (i.e., fresh or severely underburned fuel) in the spent fuel pool shall be performed prior to and after TSC loading to ensure appropriate fuel assemblies have been loaded.
Note: A qualitative visual verification that a fuel assembly has been burned shall be performed prior to or during TSC loading operations.
Note: A verification of the TSC or package fuel inventory and loading records shall be performed under a 10 CFR 71 quality assurance program prior to shipment for previously loaded (i.e., TS Cs transferred from 10 CFR 72 storage facilities) TS Cs.
Note: Fuel assemblies without visible identification shall only be loaded after quantitative measurement of the fuel assembly.
Note: Fuel assemblies may use burnup credit to demonstrate criticality acceptability. When applying burnup credit the documented fuel assembly burnup must be adjusted downward (decreased) to account for reactor record or measurement uncertainty prior to comparison to CoC minimum requirements Note: Nonfuel hardware is defined as reactor control components (RCCs), burnable poison absorber assemblies (BPAAs), guide tube plug devices (GTPDs), neutron sources/neutron source assemblies (NSAs), hafnium absorber assemblies (HFRAs),
instrument tube tie components, guide tube anchors or other similar devices, in-core instrument thimbles, steel rod inserts (used to displace water from lower section of guide tube), and components of these devices such as individual rods. All nonfuel hardware, with the exception of instrument tube tie components, guide tube anchors NAC International 7.1-12
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C or other similar devices, and steel rod inserts, may be activated during in-core operations.
Note: Up to four DFCs containing authorized PWR contents may be loaded in a TSC with a DF Basket Assembly. A DFC spacer is required to be positioned in the designated DF Basket Assembly corner locations for the shorter length DFCs. Independently, visually verify proper placement and correct orientation of each required DFC spacer.
Note: At the option of the user, install fuel assembly spacers for the axial positioning of the PWR fuel assembly types to be loaded.
Note: For fuel spacer use mandated by the CoC content conditions, verify spacer identification and install fuel spacers in each intended fuel loading location based on the fuel spacer plan prepared, which is based on the fuel assembly inventory and nonfuel hardware to be loaded. Independently, visually verify proper placement and correct orientation of each required fuel spacer.
- 5. Place the closure lid on top of the loaded TSC.
- 6. Remove the MTC with the loaded TSC from the spent fuel pool.
- 7. Insert the drain tube assembly, torque to 200 +/- 25 ft-lb (115, +/- 5 ft-lb for DF 1" drain tube)
- and remove approximately 70 gallons of water. Attach a hydrogen detector to the vent line.
Ensure that the vent line does not interfere with the operation of the weld machine. Sample the gas volume below the closure lid and observe hydrogen detector for H2 concentration prior to commencing closure lid welding operations. Monitor H2 concentration in the TSC until the root pass of the closure lid-to-shell weld is completed.
Note: Iflu concentration exceeds 2.4% prior to or during root pass welding operations, immediately stop welding operations. Evacuate the TSC gas volume or purge the gas volume with helium. Verify lli levels are <2.4% prior to restarting welding operations.
Note: In place of continuous lli monitoring, continuous gas purging of the volume below the lid may be used in concert with initial (prior to start of welding) and intermittent H2 monitoring (upon termination of gas purging and prior to re-starting welding operations).
- 8. Weld the closure lid in place and verify the adequacy of welds with root, mid-plane, and final surface visual and liquid penetrant examinations. Record results of examinations as required.
- 10. Release the pressure and drain the cavity water from the TSC by pumping or helium gas
- blowdown ensuring fuel cladding is not exposed to air.
NAC International 7.1-13
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
- 11. Vacuum dry the TSC using vacuum drying methods as follows.
Note: Ensure heat load-dependent vacuum drying time limits are not exceeded so that fuel cladding temperatures are maintained below 752°F.
- a. Connect the vacuum drying system to the vent and drain port openings.
- b. Operate the vacuum pump until a vapor pressure of< 10 torr is achieved in the TSC.
- c. Isolate the vacuum pump from the TSC and tum off the vacuum pump. Observe the vacuum gauge connected to the TSC for an increase in pressure for a minimum period of 10 minutes. If the TSC pressure is ~ 10 torr at the end of 10 minutes, the TSC is dry of free water. If pressure is > 10 torr at end of I 0 minutes, continue vacuum drying until dryness criteria are met.
- 12. Following successful completion of drying verification, evacuate TSC to a vacuum of< 3 torr and immediately backfill the TSC with the required mass of high purity helium (99.995% minimum). Disconnect drying and helium backfill system from vent and drain ports.
- 13. Install inner vent and drain port covers purging volume behind the port cover with helium gas prior to completion of welding inner port covers in place, and perform visual and liquid penetrant examinations of final weld surface. Perform leakage test of inner port cover to verify the leakage rate is :'.S 2.0 x 10-7 cm3Is, helium to a sensitivity of :'.S 1.0 x 10-7 cm 3/s, helium. Record results of examinations as required.
- 14. Install closure ring and weld the closure ring to the closure lid and TSC shell. Verify the adequacy of welds with final surface visual and liquid penetrant examinations. Record results of examinations as required.
- 15. Install outer vent and drain port covers, weld port covers in place, and perform visual and liquid penetrant examinations of final weld surface.
- 16. Install the MTC retaining device.
- 17. Decontaminate the external surface of the MTC and TSC to the limits established for the site, as required, and terminate TSC external cooling operations.
Note: The loaded TSC is now properly loaded for transfer either directly to the MAGNATRAN transport cask in accordance with CoC No. 9356 or to the MAGNASTOR concrete cask (CC) for interim storage in accordance with CoC No. 1031.
NAC International 7.1-14
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C 7.1.5 Loading the Transportable Storage Canister with GTCC Waste Contents
- 1. Visually inspect the GTCC waste TSC and GTCC waste basket liner to ensure that they are clean and free of debris.
- 2. Place the GTCC waste basket liner into the flooded reactor cavity or other designated waste loading location.
- 3. Load the authorized quantity of GTCC waste into the GTSC waste basket liner in accordance with CoC No. 9356 content limits and requirements.
- 4. Place the MTC containing the TSC into the flooded reactor cavity or other designated waste loading location.
- 5. Using the crane and slings installed on the loaded GTCC waste basket liner lift the GTCC waste basket liner and position it over the MTC/fSC.
- 8. Remove the MTC with the loaded TSC from the flooded reactor cavity or other designated waste loading location.
- 9. Insert the GTCC waste TSC drain tube assembly through the drain port opening and torque to 200 +/- 25 ft-Jb. Drain approximately 70 gallons of water from the cavity.
- 10. Weld the closure lid in place and verify the adequacy of the weld root and final weld passes with visual and liquid penetrant examinations. Record results of examinations as required.
Install the closure ring and weld the closure ring to the closure lid and TSC shell. Verify the adequacy of the welds with final surface visual and liquid penetrant examinations. Record results of examinations.
- 11. Drain the cavity water from the TSC.
- 12. Vacuum dry the TSC and verify dryness as follows:
a) Connect the vacuum drying system to the vent and drain port openings.
b) Operate the vacuum pump until a vapor pressure of< 10 torr is achieved in the TSC.
c) Isolate the vacuum pump from the TSC and turn off the vacuum pump. Observe the vacuum gauge connected to the TSC for an increase in pressure for a minimum period of 10 minutes. If the TSC pressure is ~ 10 torr at the end of 10 minutes, the TSC is dry of free water. If pressure is> 10 torr at end of 10 minutes, continue vacuum drying until dryness criteria are met .
- NAC International 7.1-15
MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C
99.9% pure) helium to 15.0 + 2, - 0 psig. Disconnect the drying and helium backfill system from the vent and drain ports.
- 14. Install inner vent and drain inner port covers, weld port covers in place, and perform visual and liquid penetrant examinations of final weld surfaces. Install outer vent and drain port covers, weld port covers in place, and perform liquid penetrant examinations of the final weld surfaces. Record results of examinations as required.
- 15. Install the TSC retaining device in the MTC.
- 16. Decontaminate the external surface of the MTC and TSC to the limits established for the site, as required.
Note: The loaded GTCC waste TSC is now properly loaded for transfer directly to the MAGNATRAN transport cask in accordance with CoC No. 9356 or for on-site storage.
NAC International 7.1-16