ML21242A183

Revision to a Response to the U.S. Nuclear Regulatory Commission Request for Additional Information for NACs Request for a Revision to Certificate of Compliance No. 93.56 for the NAC Magnatran Transportation Package

ML21242A183
Person / Time
Site:	07109356
Issue date:	08/23/2021
From:	NAC International
To:	Office of Nuclear Material Safety and Safeguards
References
ED20210134, EPID L-2021-LLA-0000
Download: ML21242A183 (87)

Text

August 2021

Revision 21 B

MAGNATRAN

(,Modular Advanced §eneration ttuclear All-purpose TRANsport)

SAFETY ANALYSIS REPORT

RAI Response Revision

NAC AINTERNATIONAL

Atlanta Corporate Headquarters : 3930 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com Enclosure 1 to ED20210134 Page 1 of I

Enclosure 1

RAI Responses

• No. 71-9356 for the MAGNATRAN Cask

Moderator Exclusion RAI Response Submittal

MAGNATRAN SAR, Revision 21B

Enclosure I to ED202 l O 134 MAGNA TRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

Enclosure 1

NACINTERNATIONAL

RESPONSE TO THE

UNITED ST ATES NUCLEAR REGULATORY COMMISSION

REQUEST FOR ADDITIONAL INFORMATION

• JULY 2021

FOR REVIEW OF THE CERTIFICATE OF COMPLIANCE NO. 9356, REVISION NO. 3

(EPID No. L-2021-LLA-0000, DOCKET NO. 71-9356)

AUGUST 2021

• Page 1 of 5 Enclosure 1 to ED20210134 MAGNA TRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

TABLE OF CONTENTS

ACCEPTANCE TESTS AND MAINTENANCE PROGRAM................................................................................... 3

• Page 2 of 5 Enclosure 1 to ED20210134 MAGNA TRAN Docket No.: 71-9356 EPID No. L-2021-LLA-OOOO

• NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION

ACCEPTANCE T;ESTS AND MAINTENANCE PROGRAM

8-1 Either revise Chapter 8 to state that all TSCs containing spent fuel (whether loaded from storage, or loaded with spent fuel on-site prior to transport) are leak tested in accordance with Section 10.1.3 of the MAGNASTOR SAR, Revision 9 (ADAMS Accession No. ML17293A085) or revise the MAGNA TRAN SAR to incorporate TSC leak testing requirements.

MAGNA TRAN packages that are transported using moderator exclusion rely on the TSC entire confinement boundary as a special design feature in addition to the MAGNA TRAN containment boundary that both together prevent a single packing error from permitting water in-leakage into the TSC and allowing contact with the fissile material, as required by 10 CFR 71.55(c).

However, it is not clear in the MAGNA TRAN SAR that TSCs, which are loaded and not placed into storage (load and go scenario), have the same leak test requirements at fabrication as TS Cs that are loaded and placed into storage under the CoC No. 1031 for the MAGNASTOR storage system. Section 8.1.4 of the MAGNATRAN SAR only includes fabrication leak test requirements for MAGNA TRAN containment boundary and does not include leak testing of the TSC shell weldment after completion of the TSC shell seam and shell to bottom plate weld, the TSC composite closure lid, and the TSC vent and drain port inner port covers and welds, which are included in the MAGNASTOR leak testing requirements.

This information is needed to determine compliance with 10 CFR 71.55(c).

NAC International Response to Thermal Evaluation RAI 8-1:

NAC has revised the requirements for the leak testing ofTSCs prior to shipment when crediting moderator exclusion. The changes to the CoC and SAR are as follows:

MAGNA TRAN proposed CoC Condition 6(d)) revised to make it clear to the shipper that the TSC must be leak tested in accordance with MAGNA TRAN SAR, Section 8.1.4.3 requirements.

"(d) For TSCs to be shipped under the moderator exclusion option of this certificate, the TSC confinement boundary shall have been leak tested in accordance with SAR, Section 8.1.4.3 leakage test requirements."

MAGNA TRAN SAR Chapter 7, Section 7.12, Step 8 a note was added to page 7.1-6 referencing new SAR Section 8.1.4.3.

"Note: For TSCs to be shipped under the moderator exclusion option, the TSC confinement boundary shall have been leak tested in accordance with SAR, Section 8.1.4.3 leakage test requirements. "

A new section has been added to MAGNA TRAN SAR Chapter 8 to describe the TSC shell weldment confinement boundary welds and the leak testing requirements. These requirements already exist in MA GNAS TOR FSAR Section 10.1.3.

• Page 3 of 5 Enclosure 1 to ED20210134 MAGNA TRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• 8.1.4.3 Leakage Tests for TSCs Shipped under Moderator Exclusion The confinement boundary is defined as the TSC shell weldment, closure lid assembly, and vent and drain po1i covers. As described in the MAGNASTOR FSAR, the confinement boundary is designed, fabricated, examined, and tested in accordance with the requirements of the ASME Code,Section III, Subsection NB, except for the code alternatives listed in the MAGNASTOR FSAR.

At the completion of the TSC shell weldment confinement boundary welds (e.g., TSC shell seam and shell to bottom plate), the TSC shell weldment shall be leakage tested. The leakage test shall be performed in accordance with the requirements and approved methods of ASME Code,Section V, Article 10, and ANSI N14.5-1997 [20] to confirm the total leakage rate (i.e., leaktight) is less than, or equal to, 1x10*7 ref. cm 3/s (air) or approximately 2x 10*7 cm 3/sec (helium). The sensitivity of the test shall be one-half of the acceptance test criteria as specified in ANSI Nl4.5-1997.

The TSC shell weldment will be closed using a test lid installed over the top of the shell and the cavity evacuated. A test envelope will be installed around the TSC enclosing all of the TSC shell confinement welds and base metal plates, and filled with 99.995% (minimum) pure helium to an acceptable test concentration. The percentage of helium gas in the test envelope shall be accounted for in the determination of the test sensitivity. A mass spectrometer leak detector (MSLD) will be used to sample the evacuated volume for helium.

If helium leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with the ASME Code,Section III, Subsection NB, NB-4450 or NB-4130, as appropriate. Following repair, the complete helium leakage test shall be re-performed to the original test acceptance criteria.

Leakage testing of the TSC shell weldment shall be performed in accordance with written and approved procedures, and the test results documented.

Based on the confinement system materials, welding requirements and inspection methods, shop helium leakage testing of the 9-inch thick closure lid is not required. However, due to the reduced thickness of the stainless steel closure lid (4-inch thick base material) of the composite closure lid assembly, and the presence of extended bolt holes for attachment of the shield plate assembly, a shop helium leakage test of the composite closure lid stainless steel plate shall be performed following fabrication. The leakage test shall be performed in accordance with the requirements and approved methods of ASME Code,Section V, Article 10, and ANSI N14.5-1997 to confirm the total leakage rate is less than, or equal to, 2 x 10*7 cm 3/s (helium). The sensitivity of the test shall be one-half of the acceptance test criteria as specified in ANSI N14.5-1997.

If leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with ASME Code,Section III, Subsection NB, NB-4130. Following repair and completion ofrequired NDE, the helium leak test shall be re-performed to the original test acceptance criteria.

Leakage testing of the composite closure lid shall be performed in accordance with written and approved procedures, and the test results documented.

In order to ensure the integrity of the vent and drain inner port cover welds, a helium leakage test of each weld is performed following welding of the inner port covers to the closure lid assembly using the evacuated envelope method, as described in ASME Code,Section V, Article I 0, and ANSI N 14.5. The leakage test is to confirm that the leakage rate for each port cover is :S 2x 10-7

• Page 4 of 5 Enclosure 1 to ED20210134 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• cm 3/s helium. Following inner p011 cover welding, a test bell is installed over the top of the port cover and the test bell volume is evacuated to a low pressure by a helium MSLD system. The minimum sensitivity of the helium MSLD shall be ::;1 x10-7 ref. cm 3/s, helium, which is one-half of the allowable leakage criteria for leaktight.

If leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with ASME Code,Section III, Subsection NB, NB-4450. Following repair, the helium leak test shall be re-performed to the original test acceptance criteria.

• Page 5 of 5 Enclosure 2 to ED20210134 Page I of I

Enclosure 2

Proposed CoC Changes

No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion RAJ Response Submittal

MAGNA TRAN SAR, Revision 21B

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• FOR RADIOACTIVE MATERIAL PACKAGES 11------r----,.--------------111

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE 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 applicable safety standards set forth in Title 10, Code of Federal Regulations, Part 71, "Packaging and Transportation of Radioactive Material."

b. This certificate does not relieve the consignor 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 IDENTIFICATION 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. CONDITIONS. "- IQ:; ;;-.**

This certificate is conditional upon fulfilling the re.~uir.§w~J:11sJr10 G~iltrtil1~(applicable, and the conditions specified below.

,'? '\\:~p-1/2,,, ~, \\~--' ~,;1 JI

5. y* ~

(a) Packaging

(1) Model No.:

• (2)

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 forging 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.

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• FOR RADIOACTIVE MATERIAL PACKAGES II~----------------.

1. a. CERTIFICATE NUMBER b. REVISION NUMBER 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 sealed 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 MAGNA TRAN 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 TSC is constructed of a stainless steel cylindrical shell, bottom-end plate, closure lid, 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 thick. The closure lid is 9 inches thick and is either a solid stainless 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 similar 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 11-----....---------------*1

1. a. CERTIFICATE 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 undamagedf~~ f~*J 1;1_s7,,e,Jl1~lies, which includes the four damaged fuel can locations. A damaged fue),c{a01m*ay~oe~late~,in,;each of the ~-!- \\ f ~~ I,.; \\ """~,i¥ It four damaged fuel can basket locations. The arrangel"()e1ts0ftubes and fuel posifi0n ~,dr the same as in the standard fuel basket, but the desigq~qf'>each of the four corner support ¥'/eldments is modified with additional structural support tg;:.idvide an enlarged position for a darn.[ged fuel can at the outermost corners of the fuet o'astet. Each damaged fuel can locati_o.Jfh($ a nominal 9.80-inch square opening. A darl],liJed fct~J;J or an undamaged fuel,/e'rnbly~g;iay be loaded in a damaged

• fuel can location~! r,~ ~~) ! lf ~,

Similar to the R:Y:gR basket, i( R 9a,~k~! co~. f 45 stain!ws steel fuel tubes that develop 87 bask~t locatiqn.. "!pW~Ps13epr'f acp BWR~gasket fuel *t*,1_ h ~-. I 5 77. h tube has a nominal 5 86

• h Ir.,llr,? ~ * ' I d *-1.. -!nc sqw=1r~~opeJil!§L~11 ~,~fir1~Ar1; rPl,~'I>".~~-_P?;,~l}in a~~i _nom1na : -In~ squa:e opening. The B~R ba~~tiNiWttt11_1!?~ri 1~fl~lh~lj~¥Lri'la n9j~1rcula!:_~Jllnder conf1gurat1on by side and corner sup~pJ! welg_;:rf.J~l~~]!ar~ RRl~1;~:-it~1 th~e>}Jt~f fuel ~es. The fuel tu~~s s_upport an enclosed neutron_*_a.. :bsoroer,,;s~heet::or:1\\*~~-~t0}to~r:~1otemQP sides of,;tt.ie fuel tube for cnt1callty control.

'(v,£p</" J"{!/ ~ i~~IJ\\1~~ 7'1"Y, '-'d Each neutro~ absq._fb~,r s~eet is:.§o~~>>i~l~J~,1s~};:~t,pf stainLf~ steel to protect the n~utron absorber dunng fuel;,~~~~1ng and to K((~~Jii1J1ipos1t10n. TRe:Aeutron absorber and stainless steel cover are secured to tt:feAuel 1,,_,1 ""\\'::> tube using weld posts disJr:ioufed across the width and along the length of the fuel tube. i/1 J__ ~.,,, *. -'\\ I\\.... 1\\..-,

The damaged fuel can confine~ thi'f~J'.rfia!Yrtal 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 (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE 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 '"_ ll':::; fR? If;: 193inches Package bodyv:dut~r:. 1aram 1eter"' (/JS;/ i~ches Impact lin;iitr;iia~eter,1,2a,mches *5+11 ~

Packag'e~IE!"~gth lJ ¢'\\, t7;P wi.tliq\\tt impact limiters 214incnes Y,1Jithi~12act limiters 322 inc'ifes11

~ \\.'.:,.. "' ~~

1 d7 \\

• The maximum gr0.ss,~:::::a weig,000 lb(J)

5.(a)(3) Drawings t:-~-~~ r~

The MAGNAT8A~.;.ti 1. Np 1:1:1.Gjtep;a ~G- * ~d;in ac~'5'rdance with NAC drawings:

In1 j I' ij [11\\,( \\ """""' /./ c~

Rt* /t l! i'l}p~ ' -...i,P 71160-500, RevJ"6P \\k~ gl{~Rftig1!_11~tion 'nspo~tfask, MAGNATRAN 71160-501, Rev~}, ~r,arTh:~iict::!\\., sk, MA<pJ:jATRAN 71160-502, Rev. 6{P' TngnsA*,-:~*-s'b-~B9qy~lMAGN~JR.AN 'Uf/J \\ ~~,, I '-J -t,..--4 71160-504, Rev. 2"~ Misc. ~.e I $~FPrans'port -vw.,_, K,l' Cas~;,IVIAGNATRAN 71160-505, Rev. 6P '" Lid Assembly, Transpof\\~~sl<, MAGNATRAN 71160-506, Rev. 1 f,_ask Cavity Space~, MA(3NATRAN 71160-511, Rev. 1 ~i=?,~rs~Ji aey~a.r;rt9r,""~iipping Configuration, Transport Cask, MAGK!Af.R~N v*"

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. 10P 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. SP Fuel Tube Assembly, MAGNASTOR - 87 BWR

____________ J NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET 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. 8P 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. 1 P Top Fuel Spacer, MAGNASTOR 71160-671, Rev. 2P Details, Neutron Absorber, Retainer, For OF [Damaged Fuel]

Corner Weldment, MAGNASTOR - 37 PWR 71160-673, Rev. 1 Damaged Fuel Can (DFC), Spacer, MAGNASTOR 71160-674, Rev. 4P OF Corner Weldment, MAGNASTOR 71160-675, Rev. 3P OF Basket Assembly, 37 Assembly PWR, MAGNASTOR 71160-681, Rev. 1 OF, Shell Weldment, TSC, MAGNASTOR 71160-684, Rev. 2 Det~il,?.J,Dif,cf/gs/y~e Lid, MAGNASTOR 71160-685, Rev. 8 '\\~fj}?[S~ ft/ssemb'.ty{f}:'.!1GNASTOR 71160-711, Rev. 1 r> ~STCC Waste Basket l!iioe!j; MAGNASTOR 71160-781, Rev. 1 ~~'dl Shell Weldment, GTCC fs~~~AGNASTOR 71160-785, Rev. 4~' GTCC TSC, Assembly, MAG~STOR

• . (b) Contents ~~---- 7,1J ~. c:-~ 0 ~

(1) Type and Forni 4 i:)f 11 Mater fa:[

(i) Undamage,c;!.J;WR as JJn. ~.if> ~ * 'wfj/Jf_f;et as~bly ~lj[i.jiJHJ!l(~

darn'agea fuel:is<l'spent nuclear fuel that does nothave any-visif feiei6'i;~&fe?'eft'h~ *.'", 1~\\r.qJ{i7Dwing tt@16ccurs in the reactor, assemblies that d6..not haYe m1gsin it,, aB_d assemblies witl'1 1

• missing rods that are replaced by solid stainless st~e'i:,*o_r. 1//,t, ~!t/\\Jj'lP f i., zirconium,c;cfi(1~h1, ~fthciit'ti{splac_(e~~'7olume equal to or greater than the original rods and a~semblies that,(/ t} ~1,;~&--~ do not contain str: ctural defects that adversely affect radiological and/or criticality s,:::if;ety and/or result ir:tJ,msupported fuel rod lengths in excess of 60 inches and that can be haFiqlecl;Jiff. ~~~l~f eJ~i:is. Undamaged PWR fuel is loaded into the short TSC, except for Combustion~E'ngin-eeriri'g (CE) 16x16 fuel assemblies, which may be loaded into either length TSC.

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 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 (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE 41 FOR RADIOACTIVE MATERIAL PACKAGES

1 a. CERTIFICATE 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 burn up 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 burnup, minimum initial average enrichment 1, and cool time 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.

~iiI~hW~:l~\\~~,~~~-~-. i*gli~1H1Jttolr:t~~~,-.~J;!ti~~~~f~~o/lm~fnn~+,~l*~*iti.al mqqeratgr, J~S(e(t:[$,idrip

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 une~:iched rod ma 1l b 1~;use1 I~ ~{~placement rod to return a fuel assembly to an undamaged cond1t1on. 1:J\\.r 1J r'u..,.,,,,;?,;,'.)i~ t,~y/ i{'t -t~ l.:.1 tf :?,.

Undamaged PWR fu~l'-lq~ emblies may contain nonf~e1'1}r'1ware. Fuel assemblies with an instrument tube tie J;gg'f~pair shall be loaded with fuel inse@a.nd/or top spacers to ensure proper spacing af"l,d'¢~URRQrt of the fuel assembly. Fuel insertsfand/or top spacers are not required when u,piQfu the;E;Jx1ei;ided fuel tube basket ~s~ tlliaiJ.op nozzle is adequately

• supported. Th~~nenfuel H'.~ -~ may be Joad~ ~omplet~0iQssembly or as individual components, ir:i~iyidual non,~ ~!:>, f~Jf-. rods_ or ~~,!!i_al-length rods/ro~lets.

'.art1al-length r0ts/rodl~ts /111tt1fm ~1u1d~ prov1ded_(gJ1de tube plug de~1~es are 1nsta_lled. No".-fb!.§1 hardwar "'R)ee~tn~e'.<J!f ~d1 ~_?ol t1/]l:e or cobalt-60 act1v1ty requirements 1~glable' ' JplJ t,~'J~onfl!~lr::hardware must meet the additional cool time re." (teutrofltcal!lsorbers), and Table 15 (for Type 1 neutron aoj3orb *-1 ;::;l{_fi

~\\J,.. t;i;;u Hafnium absorb~r;::is~ 1em 1es lu,F~}/2 f:lY *i":;:tJ ed for ~~stingho_u~e (WE) as~emblies and may have a max1m J,11,e~posure of ~@7 /MTO and m1,Js.ft0ave a minimum cool time of 16 years. Fuel assemblie~ may contai~ any number qf!~i~fadiated nonfuel solid filler fuel replacement rods. Activated ~ainless steel fuel rJ~JJlacement rods are limited to 5 steel rods per assembly, 1 assembly per basfet;sl~19-d.. ~~lJli;lifl~{steel rod exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless\\teef':rods *must be 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.

F[~\\JssefpqD~f11q~q,e~g~~)JhJB'!C:~~i'\\'D~l[MJfJ~tb}J~l~J~:,w'uJJfii§~!-tllt~1~jJiqf:1~ttj9gJ~tfm'~

retfi~i.'f~_ments;if;l.JabJe 5rorJJq~l$.;{I.P};~.safRrqWi:iate~f(;)r:Y§F3R,';(s*orfe3Ir~Ws,;yv9iqti~y~f is

• i~!at~lr;g~~~~i~i~;~!i?:i~ll!\\~if~~i§\\lii 235 U wt% enrichment over the

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.

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

~l2J~:Ja; 1~~:~g;,25, ~? iijfi~,i.n_~igy!~f" <?rlf~Kei,Ofµt~on soiic~as;5Erntf(~SA)IK_~T*J~

P~JmItted to_ be:!oa*~d 1n,a;]SC in fuel-stqragB locattsms~;J 1,; _1_2,_1!3,)18,-19, 20;,2§,c26_ or 27,:as strown_on Figure 1:l

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 assemblies. 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~tt:5itflssRlfiifsJ(assemblies * \\\\ }.(":,!) ~~j 1/. with a maximum enrichment greater than that d1ctate0:::~y,;:tlte burnup credit loaa1dg~13rve) may be loaded provided that a rod cluster control assem~\\!,~RCCA) is inserted in the asset:hJiilt~the enrichment is equal to or less than 4.05 wt. % 235 ~J~l'1d the assembly burnup is greater t~~ or equal to 12,000 MWd/MTU.

When loading undeftbl!._Ened fuel, the RCCAs must be f~leng:(b Ag-In-Cd RCCAs comprised of stainless steel cJfgJ'rod§,~ ructed with 80% Ag,;k~;>kl)~'Yanajii Cd absorber pellets and

• having an expos_ u~e equal1<;_ less ii "ij ~ K. )fQi.~ P2J) than 2_ 00,0Q,GtM~MTU. The basket must include a~sorber sheef~"'IA'.ith an e ~~ LB a~1~I>oensi!,~~l036 ~lcm,;.1i:::. Any assembli~s loaded without an RC~~ 1nse~~d~~%~~~-!V e~~~~)_p~m~p\\Qf~~1Do~~1n~ _Q~':l'?'~J?r _the_a,l}~llcaple_ -.. ~--.*.

~SJ~!!!~lY_L~c!.9JQg_prof1l§_,.,Bl'.IJ-!;JNP c;:redIf:Q~Jves,.and t1:ie:.pnt1cahty n~J?dfor R:CJ\\~*ll1SertI011 ar~

onJy applicableJ:osystems ngt crediting moderator exclusion. lnitiaJ enrichmenH.1p t:o_Q _ _wi:. o/~

3 t$, with no bu~O~~~~~~-t~i5. ermi _.B~~::. :re~~i~in::~~~:.. :or -~~c:::;*'........ "

t}gf~.'!~~£~.. ]----,1.., *.,,,:;;*----r*~.

":1/,c if /fl1

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Table 1 - PWR Hybrid Fuel Assembly Characteristics No. of Min Min Max Max Vendor Assembly Group Array Fuel Tubes Pitch OD Thick. OD Length Load Hybrid Hybrid No. of Guide Max Clad Clad Pellet Active Max Rods (See (in.) (in.) (in.) (in.) (in.) (MTU)

Note 1 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 14x14 176 5 0.5800 0.4400 0.0260 0.3805 137.0 0.4115 CE CE16H1 H1 16x16 2~§. fr *~5 F) 1(,,./7:l !Atso63 0.3820 0.0250,.__H""'.'l ff: 0.3250 150.0 0.4463 WE WE14H1 H1 14x14. "'"',119;_,/I ~- 17 0.55S0P il :{0.4900 0.0162 0.3674 145.2 0.4144

\\

WE WE15H1 H1 15x,:1§~ ~204 21 0.5630 crii:2i0, 0.0242 0.3669 144.0 0.4671 WE WE15H2 H2 15x?'~.,.-,_...:,_ _;f 204 21 0.5630 0.41'7(: ) 0.0265 0.3570 144.0 0.4469 WE WE17H1 H1 #1,zx1:zr,. 264 25 O}J:_~p

• ffe.p205 0.3232 144.0 0.4671

~>'I.

WE WE17H2 H2,Lr 1;7x17 '...r~,,-:,,:%600,0fo225 0.3088 144.0 0.4327

~?;,

Notes: _-.,

1. Combined number of guid~~fund instru.

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 2 - PWR Fuel Class Assembly Characteristics Characteristic 14x14 14x14 15x15 15x15 16x16 17x17 Fuel Class

Base Fuel Type 2 CE,SPC WE, SPC WE, SPC BW, FCF CE BW, SPC, WE, FCF Max Initial Enrichment (wt. % 235 LJ)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 Burn up 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. re,~ ID< fE ~et,~,~te 1 (lbs.) See Not~r..1, i;.§.Eae,Nofe See Note 1 See Note 1 See Note 1 ~ \\, ~- i Max Decay Heat per Fuel Location tf: \\},> = See Note 2,-¢,'11) 14;,See Note 2 See Note 2 See Note 2 (Watts) 6 See,Note 2 See Note\\2 * <:, '\\I,.(,it

.m.

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.

• ~r~;0~~r~:~~~~~~.~-Lnu;~~l~~~~~~~~~~~~~=. ~?~~~~!~~-~~[f=-~;:~~(b;n~y~~~:~~~g~~~~-~~L~i~e1:~~~t~~-~~§~~~~a~~2 fuel ana is plpcedinto damaged_ fuel cans:

6 Maximum uniform heat load per storage location.

7 TSC and maximum contents shall not exceed 104,500 pounds

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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% 23 su)

Assembly 10g Burnup = C4 x Burnup (GWd/MTU) + Cs Absorber Maximum Burnup 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 ~Jl~,-- 0.0657 2.06 0.0733 1.99

f1,t:a WE14 2.2 -~1; 0.Q,1,9!3} 1 wtt lr\\~*-* 2.©8::,, ; #QiQp72 2.21 0.0725 2.29

\\\\. \\\\f"'l-WE15 1.9 f',.,tl:c©-494 1.74 ***1i10~~. 1.72 0.0742 1.67 WE17 1.9 ~ '°l?,,, '. -.::,V0.0494 1.71 0.068'5,I-< 1.68 0.0749 1.61 BW15 1 8a,2!?:/' 0.0507 1.61 0.0687 1.59 0.0745 1.48 * ""-'?,,._

BW17 1 i!;I;:,;. "::::::--Q;0503 1.66 0.0§B)-;:J' ?'¥9;t.63 0.0733 1.59 CE14 -<? 'I 1.95 gf -;:1~~97 0.0738 1.90 ~;Y1h, 9 t1~~9 0.0727 1.90

• CE16 0.030,~2:r WE14 :::4."71 ~'1:0 0.0728 2.19 WE15 "---,1~9 ) i~$'9 0.0747 1.54 WE17 l('):9.,.'6&!t /,~~8 0.0737 1.53 BW15 1~148 ~~o. Q~{i;$?',1/,:1;;__52 0.0754 1.41 BW17 j.:;8., 0.~6Ji31" :~1~:59 0.0748 1.47 CE14 "'"'t,' 2:.,\\,. :s'."'~G}Qef66,e;J'.92 0.0729 1.87 CE16 0.027 2.l=,r"*0yo557 ""'1.92 0.0747 1.75 WE14 2.1'\\, *-*1 !())'. 06670,.__~ ~ 2.10 0.0743 2.07 WE15 1.9 o.o6zt'4.~ 1.60 0.0749 1.46 WE17 1.9 --- i,__Q.068:f' 1.54 0.0749 1.41 i,\\

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 4 - Maximum Initial Enrichment - Undamaged Fuel Configuration WE15x15 0 tional Conti urations - 20 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt% 235 U)

Number of 108 Burnup = C4 x Burnup (GWd/MTU) + Cs Assemblies Absorber Maximum Burnup 18 ::; Burnup Burnup Loaded (g/cm 2 ) Enrichment GWd/MTU < 18 GWd/MTU ::; 30 GWd/MTU > 30 (wt. % 235 U) ~__,__C_ 4_~__.'--C-s---+---'-C-4-~~'--C-s--+--'---C-4 -~--'--C-s----1

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 1.97 0.0740 1.99 33 2.15 0.0724 2.29 36 1.87 0.0741 1.84 35 0.027 1.94 0.0735 1.96 0.0730 2.21

Core Assembl NSA CE14 WE14 1.1 WE15 1.3 BW15 0.2 CE16 WE17 1.4 BW17 0.2 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 6 - Nonfuel Hardware Max Exposure and Required Cool Times Years Maximum Minimum Cool Time Years Exposure Hardware GWd/MTU WE 14x14 WE 15x15 B&W 15x15 WE 17x17 B&W 17x17 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 required 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 ~hJJ,~ th~-:5J1}s!.X_1,~um exposure and minimum cooling time limits for NSAs with absorber rod~;.af:eJ0i sarme'::;.,a~~p~e/or BPRAs.

4. Only GTPDs that do not ihcludf f{ \\?",~ £--*~u absorber, or pciisoh, rnds or water displacement rods are allowed contents.,, 'c;~ 'V ~"'ii

<1t~:J'

Hardware B&W 17x17

• BPRA 27.0 GTPD 93.3 107.8

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Figure 1 - Undamaged Fuel Basket Loading Profile!!.

(2)

(5) (6) (7)

)~,*

36 ading: remove 19 35 assem adjn ni)l)v,e, 18 '°t:'I~ -&Y'

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 credit 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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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 9-Loading Table for PWR Fuel - 23 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 s E < 2.3 5.7 5.8 6.7 6.9 6.3 6.8 6.8 2.3 s E < 2.5 5.7 5.8 6.6 6.9 6.2 6.7 6.7 2.5 s E < 2.7 5.6 5.7 6.6 6.8 6.1 6.6 6.6 2.7 s E < 2.9 5.5 5.6 6.5 6.7 6.0 6.6 6.6 2.9 s E < 3.1 5.6 5.6 6.4 6.7 6.0 6.5 6.5 3.1 s E < 3.3 5.4 5.6 6.4 6.6 6.0 6.5 6.5 3.3 s E < 3.5 5.4 5,..9 t'('~ fg:>3:c,,,, "'u 6.6 5.9 6.4 6.4 3.5 s E < 3.7 5 3 <f!'5l'§\\\\f I i5~3~ (J Jl (6/5? 5.9 6.4 6.4 3.7 s E < 3.9 5.3P'i\\:,,1'~,4. "* **?'. (; 6.2 6.5">4( 'l,:;*;,t 1 t 5.9 6.3 6.3

~

3.9 s E < 4.1 k5~3"~ 5.4 6.2 6.5 ',.;;#'5.8 6.3 6.3

'" l"'~ 6.3 4.1 s E < 4.3 ~1 5.4 6.1 6.4 ~8 6.3 4.3 s E < 4.5 5,2, 5.3 6.1 5'-'8°Q) 6.2 6.2 4.5 s E < 4.7 ~ '-Y"-'.:--- 6.1 5J/4 6.2 5.2 "--. 6.2

• 4.7 s E < 4.9 ' 5.1 _,6_..,Q 5.7 (* 6.1 6.1 E ::C:4.9 ~f6:'0) 5.7 *a~: 6.1 6.1, d/MTU Minimum lniti sepi,bly_:1 Assembly Av Enrichment ""' WE B&W wt% 23su E 6x1e._~ :J 17x17 17x17 2.1 s E < 2. *--....* - ~""

2.3 s E < 2.5 7.1 8.~ 9.1 9.1 2.5sE<2.7 ~~QY 9.0 9.0 ',,;,;::>7.0 I?'

2.7 s E < 2.9.,,{~619 ~ 11,?? !;~7.8 7.1 rrt_7~9 8.9 8.8 2.9 s E < 3.1 6:~, 7.0 8.8 8.7 3.1 s E < 3.3 6.8 '*i!ft(_0 ~ 8.6 7.7 8.6 8.6 3.3 s E < 3.5 6.7 A:;6 9t'. y-~J ;J4" *:;. -:l&, 7.7 8.6 8.6 3.5 s E < 3.7 6.7 6.9 ' ' 8.4 7.6 8.5 8.5 3.7 s E < 3.9 6.6 6.8 8.3 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 s 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 ::c:4.9 6.3 6.5 8.0 8.5 7.2 8.1 8.1 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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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 9-Loading Table for PWR Fuel - 23 kW/Package 1 (continued)

Minimum Initial 35 < Assembly Average Burnup:,:; 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears wt% 23su E 14x14 14x14 15x15 15x15 16x16 17x17 17x17 Enrichment CE WE WE B&W CE WE B&W

2.1:::; E < 2.3 2.3:::; E < 2.5 2.5:::; E < 2.7 9.7 11.9 13.5 14.7 11.6 13.7 13.7 2.7:::; E < 2.9 9.5 10.1 13.3 14.4 11.5 13.4 13.4 2.9:::; E < 3.1 9.3 9.8 13.1 14.1 11.3 13.2 13.2 3.1:::; E < 3.3 9.1 9.7 12.8 14.0 11.1 13.0 13.0 3.3:::; E < 3.5 9.0 9.§ w* 1r2t,6-13.8 10.9 12.8 12.8 k:;-,,-....1 3.5:::; E < 3.7 8.9 {~-1~ r~ 1i2~5.,,, [3::,6,, 10.8 12.7 12.6 fw,l*,>

3.7:::; E < 3.9 8 8.;l-,. ~,,-9_3 12.3 f3.5.,, __,,~ 1J 10.7 12.5 12.5 3.9:::; E < 4.1,8~i:1.j> 9.1 12.1 13.3 )~:5 12.3 12.3 4.1:::; E < 4.3 f'.1,_~~ 9.0 12.0 13.2 110*~ 12.2 12.2 -ct"""' "%,g:j}' ~

4.3:::; E < 4.5 -,,8,5 9.0 11.9 iO"'Sr 12.1 12.1 J ~jtffe 4.5:::; E < 4.7 11.8 102 /}* 12.0 12.0 8-~

• -~;_,;

• 4.7:s;E<4.9 11.7 12.0 11.9 E;::: 4.9 11.9 11.9 Minimum Initial~( d/MTU Assembly Avg}-a;\\ll wt% 23su E 7x17 17x17 Enrichment ~ WE B&W

2.1:::; E < 2.3 2.3:::; E < 2.5 2.5:::; E < 2.7 2.7:::; E < 2.9 20.0 20.0 2.9:::; E < 3.1 16.7 19.6 19.6 3.1 :s;E<3.3.1~.6 20.7 19.4 19.3 3.3:::; E < 3.5 13.1 y'1- ;! 1,s.7,,\\_, P'it.1-o,:,~ *, -~¥ '?...:it? **:.i 10.2..,,~tS:4 16.3 19.1 19.1 3.5:::; E < 3.7 12.9 13.81/"'1 Zls.6 j, 16.0 18.8 18.8 3.7:::; E < 3.9 12.7 13.7 18.3 19.9 15.8 18.7 18.6 3.9:::; E < 4.1 12.5 13.5 18.1 19.7 15.6 18.4 18.4 4.1:::; E < 4.3 12.3 13.3 17.9 19.6 15.4 18.3 18.3 4.3:::; 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:::; 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 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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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 28.7 28.6 3.3:::: E < 3.5 28.4 28.4 3.5:::: E < 3.7 28.1 28.1 3.7:::: E < 3.9 28.0 28.0 3.9::::E<4.1 27.8 27.7 4.1:::: E < 4.3 27.5 4.3:::: E < 4.5 27.4 4.5:::: E < 4.7 27.2

• 4.7:::: E < 4.9 27.0 E:?:4.9 26.9 Minimum lnit Assembly Av Enrich men E B&W wt% 23su E 17 17x17 2.1 ::::E<2.

2.3:::: E < 2.5 2.5:::: E < 2.7 '~ - - !8, 2.7:::: E < 2.9 ~§ 1 ":;r -(~ ~--* -.;::,

2.9::::E<3.1 ;;'ku v":;f~Jl... '-,~\\;;

3.1:::: E < 3.3 26.,8 *35~8 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:::: E < 3.9 25.9 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::::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 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 17 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 (continued)

Minimum Initial 55 < Assembly Average Burnup s 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 39.8 39.7 39.6 39.5 39.3 39.1 39.0 38.9

• 38.7 38.6 38.5

/Package

Max.

Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time MWd/MTU] [wt% 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 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 18 OF 42

5.(b)(1 )(i) Contents - Type and Form of Material (continued)

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 s E < 2.3 6.0 6.1 7.1 7.4 6.6 7.2 7.2 2.3 s E < 2.5 5.9 6.0 7.0 7.3 6.6 7.0 7.1 2.5 s E < 2.7 5.9 6.0 7.0 7.2 6.5 7.0 7.0 2.7 s E < 2.9 5.8 5.9 6.9 7.2 6.4 6.9 6.9 2.9 s E < 3.1 5.8 5.9 6.8 6.4 6.9 6.9 3.1 s E < 3.3 5.7 5.8 6.3 6.9 6.9 3.3 s E < 3.5 5.7 6.8 6.8 3.5 s E < 3.7 6.8 6.8 3.7 s E < 3.9 6.7 6.7 6.7 6.6 6.7 6.6 6.7 4.3 s E < 6.6 6.6

• 4.5S E<4.~"' 6.6 4.7 s E <4,~:i 6.6

'i/W Minimum lniljal /MTU E2::4.9 -* *6 6.6

Assembly AtiM En rich ment,,:",~- E B&W wt% 235 U et~ X17 17X17 2.1 s E < 2.3 2.3 s E < 2.5 10.0 10.0 2.5 s E < 2.7 9.9 9.9 2.7 s E < 2.9 9.7 9.7 2.9 s E < 3.1 8.5 9.6 9.6 3.1 s E < 3.3 7.2 7.5 9.4 8.4 9.5 9.5 3.3 s E < 3.5 7.2 7.4 9.3 10.0 8.3 9.4 9.4 3.5 s E < 3.7 7.1 7.4 9.2 9.9 8.2 9.3 9.3 3.7 s E < 3.9 7.0 7.3 9.1 9.8 8.1 9.3 9.2 3.9 s E < 4.1 7.0 7.2 9.1 9.7 8.1 9.1 9.2 4.1 SE<4.3 6.9 7.2 9.0 9.6 8.0 9.1 9.1 4.3 s E < 4.5 6.9 7.1 9.0 9.6 8.0 9.0 9.0 4.5 s E < 4.7 6.9 7.0 8.9 9.5 7.9 9.0 9.0 4.7 s E < 4.9 6.8 7.0 8.8 9.5 7.9 9.0 9.0 E 2:: 4.9 6.8 7.0 8.8 9.4 7.9 8.9 8.9

1. '-' means not allowed NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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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 :5 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears wt% 235LJ E 14X14 14X14 15x15 15x15 16X16 17X17 17X17 Enrichment CE WE WE B&W CE WE B&W

2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 10.7 11.9 15.2 16.6 13.1 15.4 15.4 2.7SE<2.9 10.5 11.2 14.9 16.2 12.9 15.2 15.1 2.9 s E < 3.1 10.3 11.0 14.7 16.0 12.6 14.8 14.8 3.1 s E < 3.3 10.1 10.8, 15.8 12.4 14.7 14.7 3.3 s E < 3.5 9.9 *~\\ \\l;i 14.4 14.5 ~;.1p,..1~11 ;

3.5 s E < 3.7 9:8" ~;;;,V;;10.4 14.1 0 14.3 14.2 '¥;,

3.7 s E < 3.9.7 10.3 13.9 :9 14.2 14.1 10.1 13.7 pwl; 14.0 14.0,,8/ d 13.6.if' 13.9 4.3 s E < 13.5.6 13.8

• 4.5 s E < 4.J'7tiiil 1.6 3.7 13.6 4.7 s E < 4r,:9f" ?';t3.6 13.6 "A\\'fJ~ \\:.c:;:>

E ~4.9 1.5 ':::;11,.6.5 13.5

~ ',., d/MTU

WE B&W 17X17

2.3 s E < 2.5 2.5 s E < 2.7 2.7SE<2.9 19.2 22.1 22.1 2.9 s E < 3.1 15.3 16.7 23.2 18.8 21.8 21.8 3.1 s E < 3.3 15.0 16.2 22.9 18.6 21.5 21.5 3.3 s E < 3.5 14.8 15.9 20.9 22.6 18.3 21.3 21.3 3.5 s E < 3.7 14.5 15.7 20.7 22.4 18.0 21.1 21.0 3.7 s E < 3.9 14.2 15.5 20.4 22.2 17.8 20.8 20.8 3.9 s E < 4.1 14.0 15.3 20.2 22.0 17.6 20.6 20.6 4.1 SE<4.3 13.9 15.0 20.0 21.8 17.5 20.5 20.4 4.3 s E < 4.5 13.7 14.8 19.8 21.6 17.3 20.3 20.3 4.5 s E < 4.7 13.6 14.7 19.7 21.5 17.1 20.1 20.1 4.7SE<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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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 13-Loading Table for High Burnup PWR Fuel - 20.9 kW/Package 1

Minimum Initial 45 < Assembly Average Burn up S 50 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt% 235 U (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 31.0 2.9$E<3.1 23.8 25.2 30.7 32.7 31.3 31.2 3.1:,; E < 3.3 23.5 - 24'7i 3:CJ 5*;* 32.5 31.0 31.0 4P ti'3/4:* :r' {I,"t; w-.:: ff; *it~4; 35.~,(f,_f3'Q 2 30.8 30.8 3.3:,; E < 3.5 23,Q \\-1' -.v>> ':,:>"'b,& *,..,:]}' {;/,".

3.5:,; E < 3.7,. 22A9" 24.1 30.0 32~ 30.6 30.5.',

3.7:,; E < 3.9 23.9 29.8 31.9 30.3 *2:>22.6 "

3.9:,; E < 4.1 23.6 29.6 31 30.1 f!tir,,-,:-,

4.1:,; E < 4~3i:t 29.4 29.9 ll :

• 4.3 :,; E <Mzt!§f 2~,Q) 29.8 4.5 :,; E i'Jr~o/ 29;,Zj 29.6

'f%.1C\\.)

4.7:,; E <"'4~9 29fo-29.5

('j E ~ 4t(~ ;:_[it 29-t:4,h 29.3

{fi(~ '

Minimum Initial ~\\IY_d/MTU Assembly 4v.9. nt~

Enrichm:nt\\ if B&W

~l~

wt % 23su E'- :iX17 17X17

2.5:,; E < 2.7 2.7:,; E < 2.9 2.9:,; E < 3.1 3.1:,; E < 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$E<3.9 28.1 30.2 35.4 37.6 36.8 36.6 3.9:,; E < 4.1 27.9 29.9 35.2 37.4 36.6 36.5 4.1:,; E < 4.3 27.6 29.7 35.1 37.3 36.4 36.3 4.3:,; E < 4.5 27.4 29.5 34.8 37.1 36.3 36.2 4.5:,; E < 4.7 27.2 29.3 34.7 37.0 36.2 36.1 4.7:,; E < 4.9 27.1 29.1 34.6 36.8 36.0 35.9 E ~4.9 26.9 28.9 34.4 36.7 35.8 35.7

• 1. '-' means not allowed NRC FORM 618 (B-2000) U.S. NUCLEAR REGULATORY COMMISSION 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

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% 235 U (E) 14X14 14X14 15x15 15x15 17X17 17X17

2.1::; E < 2.3 2.3::; E < 2.5 2.55E<2.7 2.7::; E < 2.9 2.9::; E < 3.1 3.1::; E < 3.3 3.3::; E < 3.5 42.2 42.1 41.3 41.9 41.1 41.7 35.7 41.0 41.6 5 40.8 41.5

• -!:l 41.3 41.2 41.1 41.0

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 14-Low Burnup PWR Fuel Loading Table-22 kW/Package

Max. Min. Assembly Assembly Minimum Avg. Avg. Initial Cool Time Enrichment Burnup [wt% 23sU] [Years]

[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

• 'I"))

<f' f~ 1cr*'\\ lr<H~ fi ~

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1) Contents - Type and Form of Material (continued)

(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.T~~rmeet

~~itWf~tj_o~~~;~l~t~bl~J:J~l~~-i-~~~&~~!2~~5~~.;&~~1~z~k0~~~~~i:1s:~::ti!i9:~~~; 0~1

'.A>*c *wu*,-.* A /4 AF, ~ h%***,-,.*, Jtr, 'M.., 9.,/**..,.... l.l.,;J;I,,...,,, -,~+, *,> )_

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 fy,el~~ffi]il&~ti~sfn(mlei$31flaged fuel basket, the fuel must meet the class enrichment, post-i~rapiafion cooling time, burni:lfil pit loading curves, and the TSC neutron absorber she~t}0B density in Table 16. For th l@aging profiles up to the 33, 35 and 36 assembly loading ~ttf1rn, the PWR fuel ~ust m~et the bu_rhy~,loading curves in_ Table 1_7. A short-loaded 33, ~-Q 6'r~_6... assembly loading profile may sl!Jt ufiHhe burnup credit curve in Table 1~ire~'-. ';l~~-ations 1 ~ ~rovided the /~1l for that proJt~;,i[~wn i~.:fEigure 1 are left ~mpty, at a minimum. For e~SC with ijged fuel basket "' ;..--mbly thqt.does not contain any damaged

,,}'fi,,;*:*-~y,,,,_s;c,;y;;c* <yg:"','\\.::,'""'t*~~ :~ ~~*:,,,, if tif 'i\\

fuel Y,ff - -~'.~-' th lyss~~ficmm' -irradia~i~_J;l,;:cooling time, burnup credit for all loading curve the location le <:'l!:f"stems>rid:tl',ere'diUff T Bc:f~n-~~ty4Ji!~ble?3And 4,fo"ocie"i:afo11 E~}iX ~~--~sed

~xcrU~to'.n]J ~--, y ;:,_,*_;",_;_,_;_~:+}.. /N J ->x q *.,* * + ' *,

Fuel assembly b/'ur~nup:,*JU',

• EL~nric,Sent9, an1;tcool time -"-:-,,\\/ S: requirements are provided in Table,s~~-11,: ~- 'ft\\t~:i:'¥-* neutr~~~bsorber~§l.!'!9.I~,~l~~J?-14 for baskets with _Ty~e- ~,1('1eutr absor1t. 1,,ST?S.. ;if,¥ontain_~-~~Ci:lam~ged Q!l!l~9.ti:~-~l!;!flJ]~, fuel, all fuel assemblies 1n th~~r§C must mee -"tfd1t1onal cootrt11:rfe requirements 1n Table 18 for the assembly type that is lcl~ded in the damaged fuel can.t-JJ~Mo types of fuel assemblies are loaded in different damagtd fu§I cans in a single l$C, lhe longest additional fuel cooling time applies to all fuel _assemblie~"itlffh~(~Sq:~Tbjf3aijitional cool time requirements in Table 18 apply to ass~.m911~~ l~a~~d in TSC oasl<ets with Type 1 or Type 2 neutron absorbers.

Damaged a,~~ctrfi[6~Jf~ull 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.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(ii) Contents - Type and Form of Material (continued)

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 use?}~areplacement rod to return a fuel assembly to an undamaged condition. Damaged or:l'.liqn~Jftirhup 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 1

cp,m.' P....... Pn.". e*n.*.. 1~,!;j~i*.. d*.. e.""'.".t.u~e }3,D*5. h9..r.~. C).r. ~-.ifllil§l_~... r.. _d~yic... es,... a!:.E.... **~.. t... e.. e! i1J,:>.8rt*****.*.. * *. ~... *.. ***.A*;******p*** 3.§Wt.. p.* *.. ~.* 1.*.i*C1tb1.P.... trn.1,.~.:;:g.*. ;~riator:

exc:rusio** * *.. ws.ir're1easih 'Hhefmaximum:imitialenricffmenr rcr*5*)Jvt!,%*i w 0 2butm:0**'.

~'--~_\\'.,,.'.~:<)\\~)}io:,,* */4,.,:,,"h*S *.-.*-,* A*?< )>>g/".,. ':V/12."".c... *>~,f~.s,..,72..'. -- f. *,,.;/,:(*.dy**,,>y *,,o~*o,.';'A*'\\')i.,"v**,..f *:-,-c,'IJ. *. -,.,1/*.~,c*L 'Ji,_,,p r,EtGtPJrern

Undamaged PWR fuel assemblies may contain nonfuel hardware, while damaged PWR fuel assemblies shall not, with the exceijt~on,o.f tJ;ie following unirradiated nonfuel hardware:

instrument tube tie compone@!~,fgy@\\ tb1'eta(i;ji,r§l or similar devices, and steel inserts. The nonfuel hardware may bse>l~ae'd as a complete a°'s'~ebJ.PJY or as individual components, individual nonfuel rod"S'1iii-ay be full-length rods or partiafr;lepgth rods/rodlets. Partial-length rods/rodlets are perflil1trid in guide tubes provided guide fi.Jli>'~)plug devises are installed. Fuel

~,/ ~

assemblies with an rl'lstrument tube tie rod repair shall be loa~~d with fuel inserts and/or top spacers to ensur~,.pro~1?sp;;icing and support of the,tfu~ssfii1::>1y. J '\\' \\!"':'""*,f*A'*\\:-*, v)'.:_f;;8':; _ _,,,_,,f'* ~ Fuel inserts and/or top spacers are notkhe,quired.. """" *, ~ing the extendrd;frl:~fitube bastet because the top nozzle is

• adequately sup~0rted. No ardwa~m,usV' ~;~:'fie expos!fre! and cool time or cobalt-60 activity requirer:plnts listed iis 6J"tr~ {

HFRAs are ont;~llowe. ~9i!~~,);\\~.. 1,;,;:.~Je~,~~,;mayt6;ve a maximum exposure of 4.0 GWd/MTU"~Md mti wi~rw !w~J,lime o;fA!:glears_i!iuel assemblies loaded with nonfuel hardwa~(~IJ1us.. e.,, Jt(o ~v ~<ifR~, timeJp={~uirem@~s of Table 5 (for Type 2 neutron absorbJ~}i,and"!~.'"'")~**~fWf~i R~.. }*1'.*.R~Y.!jifabsor*b*ii~t 41 Fuel assemblies may contain any number of un'iif~diatedhonlyel,. '{!!!~~:Ju '*' lace~~~1ods. Activated stainless steel fuel replacement ro~~~fe limited to,.**,,Ji,f6as asse,p;i,ely, 1 assembly per basket, and a maximum steel rod exj<tsyre of 32.5 G00d/MTW. Fue~~-~semblies with activated stainless steel rods must be cooled fo(either \\;";.,~ *~>:? a minimum of 21 years or the loading table minimum cool time (as adjusted for additional codl(tim~~JOJi',,nonf(del h;,ardware and the presence of damaged fuel in the TSC, as applicable) plus 1 year;~whfc'heV"ejr 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.

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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One NSA is permitted to be loaded in a TSC in fuel storage locations 11, 12, 13, 18, 19, 20, 25, 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%rl,p ~" i,.,..-s,-( }~ '( \\:a,,,_ and,~P/91Cd absorber pellets and having an exposure equal to or less than 200,0Q0~[';.!IYiJd!MifU.11 lihe,.ba;s~~j: must include absorber sheets with an effective 10 8 areal density~~t,l~d36 g/cm 2. Any asWemblies loaded without an RCCA inserted

\\: p "i',;.!1 '"' r---*---;--...

bred1tcarve~';,"andJtne::-errt1cal1ty ~~~ m,~~etJlJ b_~r~n~R;Y~8.i! !oa.~H~g~~~.rvet2~.thE:,,ap~i~~-lil~;as~~m?IY_lo<1/2!in@ requirernenkfor RCCA1nsert1on,7:are onls, apphcaole to systems wofil~-~Buruu~t

, *,. **. <* *. *..*. * **.. *_.. r~-*

n~! cr~R.itir~;n;io9~ratq~*ex9lu1sioQ;;,. l~itj_al enrJ.c;hrn~nt UP to.9 -wt_ ~ 2351!,.~lt

LD.Q Q!J.!!1!.HLQJi R'©CA!'re'guiremefut; isfnermitted'when;'erediting** '.mdder:ator*exclu'sion~

~=--**. (2:"-wt:'.~~;~. ~ *****=* - ~,:;.***---*-*----~.,

• '-71/C,"':;;J,;._,*.,,':Z>:,7,Z'"' hf},:;,~,.*,;..-; ".,1_-~,<'.\\~"-J~Y~- :\\_\\"_,*,\\>H:.~*,-~ *" *\\\\:,,<::,_,,.~'A~,_:-*' *JJ ~~"'-,. /~~,0 *,. \\*,,,,..,, '.'

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(ii) Contents - Type and Form of Material (continued)

Table 16-Maximum Initial Enrichment - 37 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt% 235 U)

Assembly Absorber Burnup Burnup 18::,; Burnup 30 < Burnup 50 < Burnup 10B Zero (0) = C4 x Burnup (GWd/MTU) + Cs

ID (g/cm~ Max. Enr. (GWd/MTU) <

(wt%) 18 (GWd/MTU) ::;; 30 (GWd/MTU) $ 50 (GWd/MTU)

C4 Cs C4 Cs C4 Cs C4 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. 1a"%9 **~ 1,.\\%V/S1'\\ : ~o lQi>l~i 1.74 0.0758 1.52 0.0758 1.52

~~ 'qJf \\1<

WE14 1.9 0.©542".,,1;,"/.'C':t 'i:~..,.. 'W-.;f' 1.85

• o~cr1~'?f 0.0794 1.75 0.0794 1.75

£1,. & V WE15 1.6 ~~©10482 1.43 0.0692, 0.0738 1.08 0.0738 0.767,.,

WE17 1.6 ~t,:;;:;:w-,.,._.z ff 0.0439 1.45 0.0657 1.35 1#'~0.0732

'&;1/4 A 1.00 0.0732 0.700 BW15 1.31 "0~0740 0.896 0.0740 0.614,' 4 BW17 O~Q?}i5 0.937 0.0745 0.655

• CE14 0.0:78~ 1.37 0.0781 1.37

,f "'<~

CE16 0.030 0.0%24 1.52 0.0724 1.52 WE14 0.08~ 1.50 0.0821 1.50 WE15 O.Q/$6 0.859 0.0746 0.575 WE17 ~---~ o.oz;10 0.968 0.0710 0.691 BW15 Ar'QJo/25 0.857 0.0725 0.581 BW17,;;;0~0724 0.918 0.0724 0.639 CE14 4/YJ ~< J'.:ER,0486 0.0778 1.32 0.0778 1.32 (i)/' ~i CE16 0.027 1.8 cf.049J;t 0.0761 1.33 0.0761 1.33 WE14 1.8 0.0535.; 0.0805 1.52 0.0805 1.52 WE15 1.5 0.0465 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 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(ii) Contents - Type and Form of Material (continued)

Table 17-Maximum Initial Enrichment - WE 15x15 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt% 23 su)

109 Zero (0) = C4 x Burnup (GWd/MTU) + Cs Number of Absorber Burnup Burnup 18 ~ Burnup 30 < Burnup 50 <

Assemblies (g/cm 2 ) Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Burnup (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 1.34 0.0733 1.20 0.0733 0.847 33 1.7 0.0780 1.19 0.0780 1.19 36 1.3'3;,,} 0.0731 1.02 0.0731 0.693 35 0.027 1.46.0738 1.13 0.0738 0.775 33 1.51 W784 1.09 0.0784 1.09

• .... 1,.,:ft!WR Fuel Contents Max Assembly Min. lni~!;A WE 17x17 t,.

Average Burnup AveragELl:;nri Cool Time GWd/MTU w\\.Wd-123 ears ears 35 44'1' '@, -:W-'.. 2.5 N/A

~3;, 0.8 0.3 2.~,,!i'b 3.3 2.8,.... ;;

• 1.2 0.8 40 2 7'/r; 2.9 ~'!,;;!* 0.0 0.0. ~.:.:

2.7 Jr~~.~-,~-3/4' 4.5 4.2 45 2.9 \\f.. 2.7 2.2 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 50 2.9 3.6 2.8 3.5 2.8 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 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1 )(ii) Contents - Type and Form of Material (continued)

Figure 2-Damaged Fuel Basket Loading Profile

(2)

(5) (6)

1 ih],4) (1

9) (10) (1~1): {l2j t[3:0

,,.., ;~.A

DFC designated locations may contain a loaded DFC or an undamaged PWR fuel assembly.

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5.(b)(1) Type and Form of Material (continued)

(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 of u~anjum dl9~(9E;l.~Rellets with zirconium alloy-clad fuel rods and zirconium alloy-clad water rods. 0'ft:l:roles!f1E:mpty1r,oci P,Ositions must be filled with solid, unirradiated, nonfuel filler.:,rl:>J;Js~fhat displace a voiofntLequal to, or greater than, that of the fuel rod that the filler rod r,ep"ikJes. Prior to irradiation, the t~'ifassemblies must be within the dimensions and spt~Ki~~tions of the hybrid assemblies list@)i~ Table 19. In addition, the BWR fuel must meet th.tr fDeJ.,gJass assembly specifications l~d1 il'Fillable 20. Fuel assembly burn_~p, minimum in!tial avwag~\\tQ'r:ic!:lment10, and cool time.g1Q1rem~Jih, are provided in Table 23, g__1 and -7T'ableil5J l,11:,J **.r

• '--~*-"***=--***'~~-W--*l,,.-,;:---~- l [.....-\\,'rv).

Undamaged B~f fuel mu ~""'r[ he,tfyb~i~ 1fJ~1)-c:J.fnbly enric:@,ent and the TSC neutron absorber sheet,~-~B density r *---, '*. tliryg:::u1Mt61tme~8[(:t,ifl_;lcl 8~ asse~Jy loading patterns for fuel with axial blankets in i;:11~,e. :~turl-w,it~o~ft~~~f'blapkets in.;frable 22. Spacers may be used t~ axi~11/'68sitio_nifj~11 VJ1 ltf~W~ ~9Pi#iWil~r a~i!lfbve~nt_in the T~C. Unenriched and ~rnrrad1ate~_r~el IS ig9,{3l1:J.ir~Jj.z~~lf91ili'Jl~lfl.~, ~xl'~pt ~hat y..Q,;3nnched axial. blankets are permitted, prov1d~t\\t~at t~1~rfQ1Rmal)])d@iJ~~~:\\\\t6~.:~i~ket 1s _!;;.9-~greater than 6 inches.

For a TSC that is le'$i>Nm \\~. ~ e1;/;:iA1f~~;1 fully load~tt{~~tPPtY fuel,,stora~.e~joca~io~s *i:,7:;,j.*-~~ shall begin with lo~atio~

44, followed by locat1ons;1;'.3, 45, 33, 55, 32, 56, 34, 54e~-fitd,-contmumg outward, as required, m an approximately symm°etfic ~9ttern as shown in fl~ure:3. Allowable fuel assembly locations for the 82 assembly BWR fuel ass.~m01y,:,b 9 s~et1~pnfi1g'uration are shown in Figure 4. Prior to use of the 82 assembly configuration, the.-c'ente~ ce*ir:weldment 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.

~.~e ?~-A?[~mbly9]nfigun:1.tion is the.result o_fc::riticalttY constr~Intson rjiciximwcl:.enrichrnentl YV,hen crediting moqerator exclusion this configurationiis notre_quif§q_c:1~fqll.~§1:gfc::_i!~_(§.f1 Ai2~n,blylJ~ p~rr:riLtte_c:l_~t ~o,_injtial __ enric::hmenf up tg_ §jyL_o/o ! 3_~,iJ'.

• ---JO 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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5.(b)(1 )(iii) Contents - Type and Form of Material (continued)

Table 19-BWR Hybrid Fuel Assembly Characteristics Geometry 2 *3

Assembly of Fuel of Partial Pitch Clad Clad Pellet Active Number Number Max Min Min Max Max Max Type Length OD Thick. OD Length Loading Rods Rods 1 (inch) (inch) (inch) (inch) (inch) (MTU)

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 0.4839 0.03000 0.4140 150.0 0.1841 88 618 61 N/A 0.6'.;t00/?,,.. 0.4830 "oqo~*9oq 0.4160 150.0 i91 0.4*&36;;.* ti ~ ~ *** - 1*~ p.p3000 0.4140 150.0 << 0.1872 88 62A 62 N/A,;;;:.O't64]:7:r 0.1921 88 63A 63 N/A \\: -.. t, :::,G.6420 0.4840 0.0272$.1:,. 0.4195 150.0 0.1985 88 64A 64 N/A ~=-.:.'3/4; JJ 0.6420 0.4840 0.02725~ 1 ;***~ 0.4195 150.0 0.1996 88 648 4 64 N/A""'-;" 0.6090 0.4576 0.02900.*l::.?~0;3913 150.0 0.1755 89 72A 72 N¥Pf) '£~"""."0:57:_20 0.4330 0. ~-601);;;/',;:.Q.(~740 150.0 0.1803 89 74A 74 1 1!:aJ "S\\' ""******** 0.4240 0~02~9'o 0.~60 150.0 *Q?~~Q.0,"./ 0.1873

,C/* ~*-

• 89 76A 76 {NIA o1"$Xiq %--% 1. J0_:0~1}90 01(37,50 150.0 If *r*/.: 0.1914

~' 1~~1-¥-Gl 89 79A 79 11;fj7A o}§?i}~:0,1 I 0~4240,.,'3/4*;*::--c390 0:$Jl30 150.0 0.1979

~ :t '...,,

• 89 80A 80 aJ~/A 1\\-, 0:5J£*-~',-j * --- -r10%2~0.... * ~ '--.. r-'*,950,, d 0)§.65 150.0 0.1821 810 91A 91 1 ;7;£1 ~~~9.s*1:q, 0?39 f  ;,,-, *** *}023-§/3'.,/.: 0.G*:W 150.0 0.1906 B10 92A 92 1.,,,;1*4 " a-,~~)~;1,.p'o ;J P~Q#~: 6.o~.$.:~c!>;t' Qr55 150.0 0.1946 810 96A4 96 1 ~121 ~~))4m~o ----~: o.. oi~:3'<f 0~~24 150.0 0.1787 B10 100A4 100 NJ.~;} -~Q~~B§P ~t*. *:ifoi~3o _.(~;3224 150.0 0.1861

,~;:9. u ({i,' ~ -(;;L~} lfjj '">,;? ~"-*

Notes: ",4V "' ~ ?~'. "4 (,"'"I; I.~ ;} ~-- -:,,~..,,

Assemblies may contain45§rtial-l,ength fuel rods.. '~~

2 Ass~mbly characteri~tics repr~:~~r.it,9f!~* _nominal configurations.

3 Maximum channel thickness allowed~rs l~P rry.11$ (nominal). 1~ni_r_r;~_9i~l~?j,

4 Composed of four subchannel clusters.

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5.(b)(1 )(iii) Contents - Type and Form of Material (continued)

Table 20-BWR Fuel Class Assembly Characteristics

Characteristic Fuel Class 7x7 8x8 9x9 10x10

Base Fuel Type 11 SPC,GE SPC,GE SPC, GE SPC, GE,ABB Max Initial Enrichment (wt% 235 U) 4.5 4.5 4.5 4.5 59 72 60 7413 9113 Number of Fuel Rods 49 62 76 9513, 14 48 61 9213

5412 80 63 79 10014

p >,..~~4.,-" <

Max Assembly Average Burnup (MWd/MTU) g0,00~. ~Jj§o:~ §Q;Q[~ ~f,J]~:@

Min Cool Time (years) /. ir, l;'ii iw{ t:'. ~ "\\~ /" / 4 4 4 Min Average Enrichment (wt% 235 U) 15,,,\\,/! [;~ii 1.3 "ici // ~.1,.,,~ 1.3 1.3

" <vg Max Weight (lb) per Storage Location/}"\\!.""'"'/ See Note 1 See NOD:Y1 See Note 1 See Note 1

.~, "'3/4t'tl~ C)P-Max Decay Heat (Watts) per Fuel;J,;q:§lation 253 253 253 253

-,.;, ')

Notes:

• 1. Maximum weight per _starag ** 39t~i~~i?_c space~l!nd channel) with a maximum contents weight of 64I~56 lb bly lengt~:ij 176.2 inches for BWR/4-6 assemblies and 171 Jnches the maximum nominal assembly width is r~ ---*J / f,~~{h\\iJ

2. ~i~ :scshee~bly weigJJ~clud ;~.:J::'i!

3. Maximum initial enrichment hi~nt.

4. Water rods may occup\\*rb,o -~el as~~bly to contain nominal number of water rods for the sp~~lttc.; > r/;~J

5. All enrichment values are'",notnin - ; 'rica 10n vah.,tes-~"7

0

6. Spacers may be used to a~i~]iposition fueI a'~s fnblies t~ fctfltlta); handling.

7. Each BWR fuel assembly may'l;)ave a,zirconium-based,alloy~hannel::;; 0.120 inches thick.

'j~# ~'~~~~~{

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 contains vendor information by fuel rod array. Abbreviations are as follows: General Electric/Global Nuclear Fuels (GE),

2 Exxon/Advanced Nuclear Fuels/Siemens Power Corporation (SPC).

3 May be composed of four subchannel clusters.

• Assemblies may contain partial-length fuel rods.

14 Composed of four subchannel clusters 15 Assembly average burn up is the burnup value determined by averaging the burn up over the entire fuel region (U02) of an individual assembly, including axial blankets, if present.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

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 /cm 2 Absorber 17 0.0225 10 8 /cm 2 Absorber 7 0.02 10 8 /cm 2 Type 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%

B8 59A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%

B8 60A 3.8% 4.5% ~3.7flo;i 3.5% 4.2%

!S.%=-~ *~"f\\ r.:t --;t B8 60B 3.8% 4.5% 't""ss Ii 3.6%'" 3.5% 4.2%

B8 618 3.8% "'4:~Lo 3.6% 3.5% 4.2%

<t "c;."'0 B8 62A 3.8% ~U.5% 3.6% 3.5% 4.1%

IIY\\ > 3.6% 3.4% 4.2%

B8 63A 3.8%

B8 64A 3.8% 3.6% 3.5% 4.2%

• B8 64B 3.6% 3.3% 4.0%

B9 72A 3.8% 3.4% 4.1%

B9 74A 3.7% 3.4% 4.0%

B9 76A 3.5% 3.3% 3.9%

B9 79A 3.7% 3.3% 4.0%

B9 80A 3.8% 3.5% 4.2%

B10 91A 3.7% 3.5% 4.1%

B10 92A 3.8% 3.5% 4.1%

B10 96A 3.7% 3.4% 4.0%

B10 100A 3.6% 3.4% 4.0%

• 16 Maximum planar average.

17 Borated aluminum neutron absorber sheet effective areal 10 B density.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 22-Undamaged BWR Fuel Assembly Loading Criteria Enrichment Limits for Fuel Without Axial Blankets Max. Initial Enrichment 18 (wt% 235 U)

Fuel Absorber 19 0.027 10 B /cm 2 Absorber 19 0.0225 10 B /cm 2 Absorber 19 0.02 10 B /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% 4.3% 3.5% 4.1%

B8 60B 3.7% 4.4% 3.4% 4.1%

B8 618 3.7% 3.5% 4.1%

B8 62A 3.6% 3.4% 4.1%

B8 63A 3.7% 3.4% 4.1%

B8 64A 3.7% 4.1%

• B8 64B 3.6% 4.0%

B9 72A 3.7% 4.1%

B9 74A 3.6% 4.0%

B9 76A 3.5% 3.2% 3.8%

B9 79A 3.5% 3.2% 3.9%

B9 80A 3.7% 3.5% 4.1%

B10 91A 3.7% 3.4% 4.1%

B10 92A 3.7% 3.4% 4.1%

B10 96A 3.6% 3.4% 4.0%

.... 1,,- )' (""""'"'

B10 100A 3.6% 4~3;% 3.4% 4::Qo/6 3.3% 3.9%

• -- **---** NoJe:..,\\f,Vhen:.crediting --* - -- --******** rqoder~tor exclusipn,the maxi_mumalloweclJnitial_enrichmenLis,5:wt 0/o'ttf;UJor_all......... ---*r;t*;:,...,../\\ _____.,3/4. __,J\\_,_-cf1... *....... "< ~< * > -*-*

Q~s)<~J/absQrber g_on,bjbatLods.

• 18 Maximum planar average.

19 Borated aluminum neutron absorber sheet effective areal 10 B density.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Figure 3-Undamaged Fuel Basket 87 Assembly Loading Profile

~ TSC ALIGI\\JMENT MARI-<

X = Designated NonFuel Location NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

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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)

(11) (12) (13) (14) (15) (16) (17) (18) " X l (19) (20: (21) ~22) (23) (24) (25) (26; (27) _,

(28) (29: (30) ( 31) X (33) X (35) (36) (3"7_: (38)

• (39: (40) (41) ( 42) : 43) X (45) ( 46) ( 47~ (48-) 49'

\\ J (50) ( 5l) (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: /

' (85) (86) (87)

TSC ALIGNMENT

'" MARK

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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 Burnup:;;30 GWd/MTU Assembly Avg. Minimum Coolin Time ears wt% 235LJ E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6

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:s;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 5.5 10.7 5.0 10.0 10.6 ' 'l:,i) i= -~(l*~~, 9.6 10.0 3.3:::; E < 3.5 6.0 8.8 ~ 1"" j;f5\\5 '""1t:~-q/~!,? 4.9

)'fi',trl}~ **

3.5:::; E < 3.7 6.0 8 4,_.,., 5.4,--, (;! 9.1 9.6,;. 'ttf' 9.8 <',,,,. ft/ 4.9 ~ ~

~1:c ~ 8.8 9.2 3.7:::; E < 3.9 5.9 \\s?o 5.4 9.4 ~-:4.9 3.9:s;E<4.1 5.9 7.7 5.3 9.0 ~8 8.4 8.9 4.1:::; E < 4.3 5.9 5.3 ~ 8.0 8.5 4 I;}

4.3:::; E < 4.5 5.3,,_/ 4:a~<l, 7.7 8.2

• 4.5:::; E < 4.7 4.7 e,D 7.5 7.9 4.7:::; E < 4.9 4.7 7.2 7.6 E ~4.9 4.7 t::;...~"",-.:::,' 6.9 7.4 Minimum Initial "35 GWtl/MTU Assembly Avg. ears~

wt% 235LJ Enrichment BWR/2=3 Rft.:*3u BWR/4-6 BWR/4-6 \\\\,,...:,*:,,,I ; <).: ~ ~

E 7x7:t"' "'71 ' ~p 9x,9;:;;i 9x9 10x10 2.1:::; E < 2.3 v*' ' *y tt:~c,, \\7.$,,,-....,ef,,

~ _;,-,(;J 14.3 Qft6 15.0 15.5 rf>

2.3:::; E < 2.5 8.9 \\;

& '*' f/>-

2.5:::; E < 2.7 8.8 fol§- 1'3 5 ~- 6.5 14.1 14.6 V? 1/

~ 6.4 13.4 13.9 2.7:::; E < 2.9 8.6 1'2.7 2.9:::; E < 3.1 8.5 6.3 12.7 13.2 3.1:::; E < 3.3 8.4 11.4 7.2 6.3 12.1 12.6 3.3:::; E < 3.5 8.3 10.8 7.1 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:c:;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. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 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 Burnup::: 40 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% 23su E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 2.3 s E < 2.5 2.5::; E < 2.7 14.6 16.9 12.2 18.0 10.0 17.3 17.7 2.7sE<2.9 13.3 15.8 10.7 17.0 8.7 16.3 16.7 2.9sE<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 15.2 8.4 14.6 15.0 3.3 s E < 3.5 12.6 13.9 *,C,,;~45& 8.3 13.8 14.3 ~.,: lJ 1,'!

,, ~i~~ \\iy-> 10.0 '.*j I 13.6 3.5 s E < 3.7 12.5 ~, t¢>*:--l,, 9.9 '"\\i 41'S't.'fll'" 13.8 "'"""',;, 8.2 13.2 3.7 s E < 3.9 12.4 ~<&)13.6 13.3 : tt ]:,~ > 12.6 13.0 3.9 s E < 4.1 12.2 13.5 9.8 12.7 f&,o 12.1 12.5 4.1 s E < 4.3 9.7 q7/f\\ 12.0 11.9 4.3::; E < 4.5,,*,;.I 11 7_9*:'.-"- 12.0 11.5

• 4.5 s E < 4.7 7.8 11.9 11.2 4.7sE<4.9 7.8 11.8 11.2 E ;;:4.9 7.7 ~t 11.8 11.1 Minimum Initial 5 GWa/MTU Assembly Avg.

Enrichment BWR/4-6 BWR/4-6 wt% 235 U E 9x9 10x10 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7sE<2.9 22.3 17.4 22.7 21.5 2.9sE<3.1 19.7 14.8 20.0 19.4 3.1 s E < 3.3 18.9 20.5 19.1 12.3 18.8 18.2 3.3 s 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 3.7sE<3.9 18.2 19.9 14.7 18.5 11.5 18.0 17.1 3.9sE<4.1 18.1 19.6 14.6 18.2 11.4 17.9 16.9 4.1 s 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 s 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 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 38 OF 42

Minimum Initial 45 < Assembly Average Burnup :5: 50 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 ~ 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 :5: 55 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% 23SU (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 - - - - - - -

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 NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 39 OF 42

Table 24--Loading Table for BWR:FueI..::. 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% 23su 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 3.1 ::; E < 3.3 3.3::; E < 3.5 42.3 44.6 40.0 43.4 38.2 43.5 42.3 3.5::; E < 3.7 39.8 42.3 37.7 41.0 35.7 41.1 39.8 3.7::; E < 3.9 38.3 40.0 35.3 38.7 33.3 38.7 37.5 3.9::; E < 4.1 38.1 40.2 33.7 38.3 30.9 38.0 36.5 4.1::; E < 4.3 37.8 40.0 33.6 38.2 29.0 37.8 36.4 4.3::; E < 4.5 37.8 39.8 33.4 38.2 28.9 37.8 36.4 4.5::; E < 4.7 37.7 39.6 33.2 38.0 28.7 37.7 36.2

• 4.7::; 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 allo,wed r-tj) ~

'~\\ ~

\\l::#P,. <<.,4::..z?"~

f;ibiii2S.----E'~wsuinup B.\\111 ~ ~oadinll~ble - ~W/Package

• • * * -"-** v;_Ap ?//1i}. ;;,r,,;fj 'l,11\\.R\\~ ~~ "'-..ri,rj ~

Max.

Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time

[MWd/MTU] [wt% 235 U] [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 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE 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 ~ont~nts d~scribed in l!ePJ. t~b)<,1',(jyt i~p ioA 37 ~ndamag~d PWR !uel assemblies, including nonfweli~,,4rciware 'and"'s 1pac~JS,t-s,w1th a maximum weight of 62,160 _pounds and a m~om decay heat limit per>?fl!E~rlocation not to exceed the values rn Table 2. '":) * ')

(ii) For the ~ontent~ J!~f r;j~~?in Item 5.(b)(1)(ii): Upr"tQ~~i~t~&Jed PWR fue_l _

assemblies, wh1pt1may rn.¢~~._up to 4 damaged ** * ** JJuel assemblies rn

• damaged fuel qails, nonfu~ "'"' 'w,are a paq, a maxiJJiym weight of 61,184 pounds (TSC clh'tJ* maxi mu e:~ts "ot.e 104,500 1rp~unds) and a maximum decay heat limTf:Jger fuel lo., t t d'

• es in Ta~Je)2.

(iii) For the conten\\;:9esc"tt~i i. 19:t{~ed $R fuel assemblies, including chann'els an.!"'. e* .'*Qf 62,@56' pounds and a maximum deca~t~~at Ii le 2 ~

1 (iv) For the contents'i~~cribe J in it e:pj.J;g' wastt~~ith a maximum weight per package of 55,000*pt~ ds in total.,. 1mum"'Elecay 1h~~at for the GTCC waste is 1.7 kW per package. ;,:~,,.; *<s~++ifr

5.(c) Criticality Safety Index

Undamaged PWR and BWR Fuel 0.00

Damaged PWR Fuel 100.00

6. In addition to the requirements of Subpart G of 10 CFR Part 71:

(a) The package must be prepared for shipment 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) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE 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).

~c;)~~~--_F:9r T~~s to~~e- ~?lpped un1~r t~tmosrrat9rex,(;,l,usior op~1w1 of}his c~rtificaJe'<only T~C,~~:th~tt are within th err 1rnt1al term for sto(age or are new and ljave;,J_!;>_~~!l.Lqa.c:i~_d *.s1nc;l_1iac;~QJDJQ storage.are authorized for u*s,e urn:lerJljqderc;1tor.~xclusion.i

(q). :.~---. Fqr TS Cs to-=-be sliipp~<;l und~r the m'oderatot exil.usioi:i.opffoi, Oftfhis :c~rtificate; the '[sq

\\ t} -,, *.,f,, J~~, f<_icl* >>"..:_~. "' 1/4_/, '<!>',"..,~'i,. :'. <:-c::,.,.1,,,0£ _.,/,-'\\* '.\\,*"',,*<"':* -*,, '.*.,,, ",,,s.,:'-,,','<: """",;c*-~

confinement_bounoary shallJiave bee'riJeaK*tested* in *accordance'!with SAR;. Section :_8.1.4:3

,,, > !~'04..-,*....___.._ _____ ~ ~ ---- ~--*~~,-~---' -*-'"" -------~~-*<"

lef;:1kage;test requJremeajs.:

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 N_a_ tional,Gar:go,.,Bureau, Inc., must have evaluated and approved r,,_ I'-'{ *XI Ir" rt

the system used to support and s~cur~ tj;j*erpackage,f0ilj~~rrge or vessel, and must have certified that Guard. '\\, package stowage is in accorda~eri'witfi the ~..., *(j regulations ofth~ C?pmmandant, United States Coast

~, 1

~'?c§C

. t::J2i;.;

Transport by air is not authdrif'._~d.

t~ ~~~

Transport of fuel assenfl:>lr~s, as\\~tti: "'

not authorized. ff"'!:"

11.

12.

13. The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 71.17.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE 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'@2embec:J1,_202fil

FOR THE U.S. NUCLEAR REGULATORY COMMISSION

.Date: ______ _

Enclosure 3 to ED20210134 Page 1 of2

Enclosure 3

List of SAR Changes

No. 71-9356 for the MAGNATRAN Cask

• Moderator Exclusion RAI Response Submittal

MAGNATRAN SAR, Revision 21B

Enclosure 3 to ED202 l O 134 Page 2 of2

• List of Changes for the MAGNA TRAN SAR, Revision 21B

Note: The List of Effective Pages and the Chapter Table of Contents, List of Figures and List of Tables have been revised accordin I to reflect the list of chan es detailed below, if needed.

Chapter 1 No Changes Chapter 2 No Changes Chapter 3 No Changes Chapter 4 No Changes Chapter 5 No Changes Chapter 6 No Changes

• Chapter 7 Page 7.1-6 Added "Note" text following Step 8 in Section 7.1-2 where indicated.

Page 7.1-7 Text flow changes.

Chapter 8 Page 8.1-7 thru 8.1-8 Added new Section 8.1.4.3, "Leakage Tests for TSCs Shipped under Moderator Exclusion" Page 8.1-9 thru 8.1-30 Text flow changes.

Enclosure 4 to ED202 l O 134 Page 1 of 1

Enclosure 4

LOEP and SAR Page Changes

• No. 71-9356 for the MAGNATRAN Cask

Moderator Exclusion RAI Response Submittal

MAGNATRAN SAR, Revision 21B

MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• List of Effective Pages

Chapter 1 Page 2.6.2-1 thru 2.6.2-10.................. Revision 0 Page 1-i thru 1-ii................................. Revision 0 Page 2.6.3-1........................................ Revision O Page 1-iii........................................ Revision 20C Page 2.6.4-1 thru 2.6.4-2.................... Revision 0 Page 1-1.............................................. Revision 0 Page 2.6.5-1 thru 2.6.5-5.................... Revision O Page 1.1-1 thru 1.1-8........................... Revision 0 Page 2.6.6-1........................................ Revision 0 Page 1.2-1 thru 1.2-5........................... Revision 0 Page 2.6.7-1........................................ Revision 0 Page 1.3-1 thru 1.3-7........................... Revision 0 Page 2.6.7.1-1 thru 2.6.7.1-9.............. Revision 0 Page 1.3-8 thru 1.3-11.................... Revision 21A Page 2.6.7.2-1 thru 2.6.7.2-5.............. Revision 0 Page 1.3-12.................................... Revision 20C Page 2.6.7.3-1 thru 2.6.7.3-9.............. Revision 0 Page 1.3-13 thru 1.3-20....................... Revision 0 Page 2.6.7.4-1..................................... Revision 0 Page 1.3-21 thru 1.3-32.................. Revision 20C Page 2.6.7.5-1 thru 2.6.7.5-34............ Revision O Page 1.3-33 thru 1.3-37....................... Revision 0 Page 2.6.7.6-1 thru 2.6.7.6-5.............. Revision 0 Page 1.3-38.................................... Revision 20C Page 2.6.7.7-1 thru 2.6.7.7-19............ Revision 0 Page 1.3-39 thru 1.3-43....................... Revision 0 Page 2.6.8-1........................................ Revision 0 Page 1.3-44.................................... Revision 20C Page 2.6.9-1........................................ Revision 0 Page 1.3-45 thru 1.3-46....................... Revision 0 Page 2.6.10-1...................................... Revision 0 Page 1.3-47.................................... Revision 20C Page 2.6.11-1 thru 2.6.11-5................ Revision 0 Page 1.3-48......................................... Revision 0 Page 2.6.12-1 thru 2.6.12-2................ Revision 0 Page 1.3-49 thru 1.3-50.................. Revision 20C Page 2.6.12.1-1................................... Revision 0 Page 1.3-51......................................... Revision 0 Page 2.6.12.2-1 thru 2.6.12.2-7.......... Revision O Page 1.4-1 thru 1.4-4........................... Revision 0 Page 2.6.12-3-1 thru 2.6.12-3-4.......... Revision 0 Page 2.6.12.4-1 thru 2.6.12.4-12........ Revision 0 36 drawings (see Section 1.4.3) Page 2.6.12.5-1 thru 2.6.12.5-6.......... Revision 0

• Page 2.6.12.6-1 thru 2.6.12.6-4.......... Revision 0 Chapter 2 Page 2.6.12-7-1 thru 2.6.12-7-3.......... Revision 0 Page 2-i............................................... Revision 0 Page 2.6.12.8-1 thru 2.6.12.8-10........ Revision 0 Page 2-ii......................................... Revision 20C Page 2.6.12.9-1 thru 2.6.12.9-6.......... Revision 0 Page 2-iii thru 2-xxiii.......................... Revision 0 Page 2.6.12.10-1................................. Revision 0 Page 2-xxiv.................................... Revision 20C Page 2.6.12.11-1................................. Revision 0 Page 2-1.............................................. Revision 0 Page 2.6.12.12-1 thru Page 2.1-1........................................... Revision 0 2.6.12.12-3............................. Revision 0 Page 2.1.1-1 thru 2.1.1-6..................... Revision 0 Page 2.6.12.13-1 thru Page 2.1.2-1 thru 2.1.2-11................... Revision 0 2.6.12.13-3............................. Revision 0 Page 2.1.3-1 thru 2.1.3-2..................... Revision 0 Page 2.6.12-14-1 thru Page 2.1.4-1 thru 2.1.4-5..................... Revision 0 2.6.12.14-2............................. Revision 0 Page 2.2-1........................................... Revision 0 Page 2.6.13-1 thru 2.6.13-2................ Revision 0 Page 2.2.1-1 thru 2.2.1-17................... Revision 0 Page 2.6.13.1-1 thru 2.6.13.1-2.......... Revision 0 Page 2.2.2-1 thru 2.2.2-9..................... Revision 0 Page 2.6.13.2-1 thru 2.6.13.2-13........ Revision 0 Page 2.2.3-1........................................ Revision 0 Page 2.6.13.3-1 thru 2.6.13.3-3.......... Revision 0 Page 2.3-1........................................... Revision 0 Page 2.6.13.4-1 thru 2.6.13.4-14........ Revision 0 Page 2.3.1-1........................................ Revision 0 Page 2.6.13.5-1 thru 2.6.13.5-2.......... Revision 0 Page 2.3.2-1........................................ Revision 0 Page 2.6.13.6-1 thru 2.6.13.6-5.......... Revision 0 Page 2.4-1 thru 2.4-2........................... Revision 0 Page 2.6.13.7-1 thru 2.6.13.7-4.......... Revision 0 Page 2.5-1........................................... Revision 0 Page 2.6.14-1...................................... Revision 0 Page 2.5.1-1 thru 2.5.1-12................... Revision 0 Page 2.6.14.1-1 thru 2.6.14.1-8.......... Revision O Page 2.5.2-1 thru 2.5.2-18................... Revision 0 Page 2.6.14.2-1 thru 2.6.14.2-2.......... Revision 0 Page 2.6-1........................................... Revision 0 Page 2.6.14.3-1 thru 2.6.14.3-14........ Revision 0 Page 2.6.1-1 thru 2.6.1-15................... Revision 0 Page 2.6.14.4-1 thru 2.6.14.4-3.......... Revision 0 Page 1 of 4 MAGNA TRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

List of Effective Pages (cont'd)

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Page 4 of 4 MAGNATRAN Transport Cask SAR April 2019 Docket No. 71-9356 Revision 0

• 7.1.2 Loading of Contents Note: The MAGNA TRAN 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.

I. Install appropriate work platforms, scaffolding or lifts to allow access to the top of the MAGNA TRAN 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 MAGNATRAN transp011 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 August 2021 Docket No. 71-9356 Revision 21 B

72 normal and off-no1mal, accident and natural phenomena events, retains its ability

• to satisfy functional and performance requirements of the MAGNA TRAN 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 MAGNA TRAN packaging certified content conditions are validated.

TS Cs 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 MAGNA TRAN 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 MAGNA TRAN transport cask containment boundary for off-site transport.

Note: For TSCs to be shipped under the moderator exclusion option, the TSC confinement boundary shall have been leak tested in accordance with SAR, Section 8.1.4.3 leakage test requirements.

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.

NAC International 7.1-6

• MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• 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.

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 MAGNATRAN 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 MAGNA TRAN 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 MAGNA TRAN 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

• MAGNA TRAN 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 MAGNA TRAN 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.

15. Disengage the MTC lift yoke from the MTC trunnions and move the lift yoke from the area.

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 of the 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 August 2021 Docket No. 71-9356 Revision 21 B

• Chapter 8 Acceptance Tests and Maintenance Program

Table of Contents

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM...................................... 8-1 8.1 Acceptance Tests.......................................................................................................... 8.1-1 8.1.1 Visual Inspections and Measurements.............................................................. 8.1-1 8.1.2 Welding and Weld Examinations..................................................................... 8.1-2 8.1.3 Structural and Pressure Tests............................................................................ 8.1-3 8.1.4 Leakage Tests.................................................................................................... 8.1-5 8.1.5 Component and Material Tests......................................................................... 8.1-8 8.1.6 Shielding Tests................................................................................................ 8.1-24 8.1.7 Thermal Acceptance Test............................................................................... 8.1-25 8.1.8 Miscellaneous Tests........................................................................................ 8.1-30 8.1.9 Packaging Identification................................................................................. 8.1-30 8.2 Maintenance Program................................................................................................... 8.2-1 8.2.1 Structural and Pressure Tests............................................................................ 8.2-1 8.2.2 Leakage Tests.................................................................................................... 8.2-2 8.2.3 Component and Material Tests......................................................................... 8.2-6 8.2.4 Thermal Test..................................................................................................... 8.2-7

• 8.2.5 Miscellaneous Tests.......................................................................................... 8.2-7 8.3 References..................................................................................................................... 8.3-1

NAC International 8-i MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B List of Figures Figure 8.1-1 Thermal Test Arrangement............................................................................. 8.1-29

• List of Tables

Table 8.1-1 Neutron Absorber Material Minimum 10B Loading....................................... 8.1-23 Table 8.1-2 Mechanical Properties of Neutron Absorber.................................................. 8.1-23 Table 8.2-1 MAGNA TRAN Maintenance Schedule........................................................... 8.2-5

NAC International 8-ii MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• other corrective actions, shall be taken to repair any detected leaks. The component and replaced seal shall then be retested and re-inspected in accordance with the original test requirements and acceptance criteria prior to final acceptance. After successful completion of the leakage tests, the quick-disconnect is installed* in the cask lid port opening and torqued.

8.1.4.3 Leakage Tests for TSCs Shipped under Moderator Exclusion

The confinement boundary is defined as the TSC shell weldment, closure lid assembly, and vent and drain port covers. As described in the MAGNASTOR FSAR, the confinement boundary is designed, fabricated, examined, and tested in accordance with the requirements of the ASME Code,Section III, Subsection NB, except for the code alternatives listed in the MAGNASTOR FSAR.

At the completion of the TSC shell weldment confinement boundary welds (e.g., TSC shell seam and shell to bottom plate), the TSC shell weldment shall be leakage tested. The leakage test shall be performed in accordance with the requirements and approved methods of ASME Code,Section V, Article 10, and ANSI N14.5-1997 [20] to confirm the total leakage rate (i.e.,

leaktight) is less than, or equal to, 1x10-7 ref. cm3/s (air) or approximately 2x 10-7 cm3/sec (helium). The sensitivity of the test shall be one-half of the acceptance test criteria as specified in ANSI N14.5-1997.

• The TSC shell weldment will be closed using a test lid installed over the top of the shell and the cavity evacuated. A test envelope will be installed around the TSC enclosing all of the TSC shell confinement welds and base metal plates, and filled with 99.995% (minimum) pure helium to an acceptable test concentration. The percentage of helium gas in the test envelope shall be accounted for in the determination of the test sensitivity. A mass spectrometer leak detector (MSLD) will be used to sample the evacuated volume for helium.

If helium leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with the ASME Code,Section III, Subsection NB, NB-4450 or NB-4130, as appropriate. Following repair, the complete helium leakage test shall be re-performed to the original test acceptance criteria.

Leakage testing of the TSC shell weldment shall be performed in accordance with written and approved procedures, and the test results documented.

Based on the confinement system materials, welding requirements and inspection methods, shop helium leakage testing of the 9-inch thick closure lid is not required. However, due to the reduced thickness of the stainless steel closure lid (4-inch thick base material) of the composite closure lid assembly, and the presence of extended bolt holes for attachment of the shield plate assembly, a shop helium leakage test of the composite closure lid stainless steel plate shall be NAC International 8.1-7 MAGNATRAN Transport Cask\\SAR August 2021 Docket No. 71-9356 Revision 21 B

performed following fabrication. The leakage test shall be performed in accordance with the

• requirements and approved methods of ASME Code,Section V, Article 10, and ANSI N14.5-1997 to confirm the total leakage rate is less than, or equal to, 2 x 10-7 cm3/s (helium). The sensitivity of the test shall be one-half of the acceptance test criteria as specified in ANSI N14.5-1997.

If leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with ASME Code,Section III, Subsection NB, NB-4130. Following repair and completion of required NDE, the helium leak test shall be re-performed to the original test acceptance criteria.

Leakage testing of the composite closure lid shall be performed in accordance with written and approved procedures, and the test results documented.

In order to ensure the integrity of the vent and drain inner port cover welds, a helium leakage test of each weld is performed following welding of the inner port covers to the closure lid assembly using the evacuated envelope method, as described in ASME Code,Section V, Article 10, and ANSI N14.5. The leakage test is to confirm that the leakage rate for each port cover is :'.S 2xl0-7 cm3/s helium. Following inner port cover welding, a test bell is installed over the top of the port cover and the test bell volume is evacuated to a low pressure by a helium MSLD system. The minimum sensitivity of the helium MSLD shall be :Sl x 10-7 ref. cm3/s, helium, which is one-half of the allowable leakage criteria for leaktight.

• If leakage is detected, the area of leakage shall be identified, repaired and re-examined in accordance with ASME Code,Section III, Subsection NB, NB-4450. Following repair, the helium leak test shall be re-performed to the original test acceptance criteria.

8.1.5 Component and Material Tests

Individual MAGNA TRAN transport cask components shall be tested, as defined herein, to ensure that the component meets the design requirements for its intended function during normal transport, loading and handling operations of the cask system. Component test acceptance criteria are established on the basis of the function, the corresponding graded quality category, and the design requirements of the component.

8.1.5.1 Valves, Rupture Discs and Fluid Transport Devices The MAGNA TRAN transport cask containment boundary does not include any valves or rupture discs. Operational access to the cask cavity for cavity evacuation, backfilling and sampling is provided by a quick-disconnect valved nipple installed in the cask lid port. The quick-disconnect serves as a valve, when the mating parts are connected, and is used to connect ancillary NAC International 8.1-8 MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• equipment to the cask cavity. The quick-disconnect is not included in the MAGNATRAN transport cask containment boundary, and therefore, no credit is taken for any containment function provided by the quick-disconnect.

Prior to transport, the cask lid port containing the quick-disconnect is sealed by the installation of a bolted coverplate fitted with two concentric O-ring seals. The port coverplate and the inner metallic O-ring seal are defined as the containment boundary. A new metallic O-ring seal is installed and helium leak tested prior to each shipment.

The neutron shield assemblies are bolted to the MAGNA TRAN transport cask body outer shell.

There is no penetration of the cask containment boundary, even during a fire accident.

Expansion foam within each neutron shield assembly provides for pressure reduction in case of a vapor pressure increase resulting from the neutron shield material being exposed to a transport thermal accident condition. The neutron shield design minimizes post-accident recovery efforts as the neutron shield assemblies remain intact in the event of an overpressure condition.

The loaded TSC does not include any rupture discs or fluid transport devices. The TSC closure lid vent and drain ports are each closed by redundant port covers welded to the lid. The quick disconnect valved nipples, which are recessed into ports in the closure lid, are used during TSC loading activities. The redundant welded port covers provide a redundant containment boundary

• at the vent and drain ports. No containment credit is taken for the quick-disconnects.

8.1.5.2 Gaskets The transport cask lid and lid port coverplate are each sealed using concentric sets of two O rings. The inner metallic O-ring provides the containment boundary seal. The outer EPDM O-ring forms an annulus with the inner metallic seal and allows the testing of the inner O-ring containment boundary seal.

The inner metallic O-rings, which are part of the primary containment boundary, are tested in accordance with the requirements of ANSI N14.5-1997. The cask lid and lid port coverplate metallic O-rings are replaced and helium leak tested in accordance with ANSI N14.5-1997 prior to each loaded cask shipment. The outer EPDM O-rings are replaced annually, or as required based on visual inspections performed during cask operations (Chapter 7).

The transport cask lid and lid port coverplate test plugs are provided with stainless steel boss seals. The plugs and their seals prevent entry of foreign material or fluids into the seal annulus during transport operations. The plugs and seals do not provide any containment boundary functions.

• The confinement boundary provided by the welded TSC has no mechanical seals or gaskets.

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8.1.5.3 Neutron Absorber Tests

• Neutron absorber materials are included in the design and fabrication of the MAGNASTOR fuel basket assemblies to assist in the control ofreactivity, as described in Chapter 6. The basket assemblies support the spent fuel contents in a MAGNASTOR transportable storage canister (TSC), which is transported in a MAGNA TRAN transport cask. Criticality safety is dependent upon the neutron absorber material remaining fixed in position on the fuel tubes and containing the required amount of uniformly distributed boron. A neutron absorber material can be a composite of fine particles in a metal matrix or an alloy of boron compounds with aluminum.

Fine particles of boron or boron-carbide that are uniformly distributed are required to obtain the best neutron absorption. Three types of neutron absorber materials are commonly used in spent fuel storage and transport cask fuel baskets: Bora] (registered trademark), borated metal matrix composites (MMC), and borated aluminum alloy. The fabrication of the neutron absorber material is controlled to provide a uniform boron carbide distribution and the specified 10B areal density.

8.1.5.3.1 Design/Performance Requirements The MAGNASTOR fuel basket assemblies utilize sheets of neutron absorber material that are attached to the sides of the spent fuel storage locations in the fuel baskets. The materials and dimensions of the neutron absorber sheets are defined on license drawings 71160-571 and 71160-572. The material is called out as a metallic composite (includes borated aluminum alloy,

• borated MMC, and Baral, which are available under various commercial trade names).

Incorporating optional neutron absorber materials in the design provides fabrication flexibility for the use of the most economical and available neutron absorber material that meets the critical characteristics necessary to assure criticality safety. The critical design characteristics of the neutron absorber material are:

• A minimum "effective" areal density of 0.036, 0.030 or 0.027 g/cm2 10B for the PWR basket and 0.027, 0.0225 or 0.020 g/cm2 10B for the BWR basket

• A uniform distribution of boron carbide

• A yield strength greater than or equal to that used in Section 8.1.5.3. 8 and Section 8.1.5.3.9

• An effective thermal conductivity greater than or equal to that used in Section 8.1.5.3.8 The required minimum actual 10B loading in a neutron absorber sheet is determined based on the effectiveness of the material, i.e., 75% for Baral and 90% for borated aluminum alloys and for borated metal matrix composites. Testing will be used to verify the areal density and the uniform distribution of 10B in the neutron absorber materials. Table 8.1-1 presents a tabulation of the types of neutron absorber materials, the required minimum effective areal density of 10B, and the required minimum as-fabricated areal density of 10B. The analyses of the fuel baskets do

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• not consider the yield strength of the neutron absorber material other than that it be sufficient to maintain its form, i.e., at least equivalent to the yield strength listed in Table 8.1-2.

The positions of the neutron absorber sheets with their attachments and retainers to the fuel tubes are shown on license drawings 71160-551 and 71160-591. The attachments and retainers ensure that the neutron absorber remains in place for all loading conditions for the lifetime of the canister.

8.1.5.3.2 Terminology Applicable terminology definitions for the neutron absorber materials:

acceptance - tests conducted to determine whether a specific production lot meets selected material properties and characteristics, or both, so that the lot can be accepted for commercial use.

areal density - for sheets with flat parallel surfaces, the density of the neutron absorber times the thickness of the material.

designer-the organization responsible for the design or the license holder for the dry cask storage system or transport packaging. The designer is usually the purchaser of the neutron absorber material, either directly or indirectly

• (through a fabrication subcontractor).

lot-a quantity of a product or material accumulated under conditions that are considered uniform for sampling purposes.

neutron absorber - a nuclide that has a large thermal or epithermal neutron absorption cross-section, or both.

neutron absorber material - a compound, alloy, composite or other material that contains a neutron absorber.

neutron attenuation test - a process in which a material is placed in a thermal neutron beam, and the number of neutrons transmitted through the material in a specified period of time is counted. The observed neutron counting rate may be converted to areal density by performing the same test on a series of calibration standards.

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neutron cross-section - a measure of the probability that a neutron will interact

• with a nucleus; a function of the neutron energy and the structure of the interacting nucleus.

packaging-in transport of radioactive material, the assembly of components necessary to enclose the radioactive contents completely.

qualification - the process of evaluating and testing, or both, a material produced by a specific manufacturing process to demonstrate uniformity and durability for a specific application.

8.1.5.3.3 Inspections After manufacturing, each sheet of neutron absorber material will be visually and dimensionally inspected for damage, embedded foreign material, and dimensional compliance. The neutron absorber sheets are intended to be defect/damage free, but limited defects/damages are acceptable. Allowed defects are discussed in each material specification section that follows.

Standard industrial inspections will be performed on the neutron absorber sheets to verify the acceptability of physical characteristics such as dimensions, flatness, straightness, tensile properties (if structural considerations are applicable) or other mechanical properties as

• appropriate, surface quality and finish. Inspection and testing of the neutron absorber materials will be performed in accordance with written procedures, by appropriately certified personnel, and the inspection and test results will be documented.

8.1.5.3.4 Specification

Three types of neutron absorber materials are permitted to augment criticality control in the MAGNASTOR fuel baskets - (1) Bora], a clad composite of aluminum and boron carbide, as specified in Section 8.1.5.3.5; (2) borated metal matrix composites (MMC), as specified in Section 8.1.5.3.6; and (3) borated aluminum alloy, as specified in Section 8.1.5.3.7. The required minimum "effective" areal density of 10B in a neutron absorber is defined in Chapter 6 of this SAR and is based on the fuel basket geometry and on the fuel assembly type and reactivity. Environmental conditions encountered by the neutron absorber material may include:

• Immersion in water with the associated chemical, temperature and pressure concerns

• Dissimilar materials

• Gamma and neutron radiation fluence

• Dry heat-up rates

• Maximum temperatures NAC International 8.1-12 MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• Except for materials for which validation has been completed, the durability of the neutron absorber materials is validated to demonstrate the following results:

• Neutron absorber materials will not incur significant damage due to the pressure, temperature, radiation, or corrosion environments that may be present in the loading and storage of spent fuel;

• Aluminum and boron carbide do not react with each other in the range of the maximum temperatures present in the fuel baskets;

• There are no significant changes_ in mechanical properties of the neutron absorber materials due to the fast neutron fluences experienced in spent fuel storage;

• General corrosion does not have time to affect the integrity of the neutron absorber material due to the very short time of immersion in spent fuel pool water.

Individual material types and process lots are tested to verify the presence, uniform distribution and minimum areal density (effectiveness) of 10B specific to each type of neutron absorber material.

All neutron absorber materials are procured and qualified under a Quality Assurance/Quality Control program in conformance with the requirements of 10 CFR 72, Subpart G.

8.1.5.3.5 Boral Boral is a composite core of blended boron carbide and aluminum powders between outer layers

• of aluminum. The core is slightly porous. Sheets of Boral are formed and mechanically bonded by hot-rolling ingots of the core material between aluminum sheets. Boral is credited with an effectiveness of 75% of the specified minimum areal density of 10B based on testing of the material. The Boral neutron absorber tests are described in Section 8.1.5.3.9.

Visual inspections of the Boral sheets will verify the presence of a full core and will identify any cladding damage, cracks or discontinuities, embedded foreign material, or peeled cladding.

Evidence of less than a full core, embedded foreign material, cracks or sharp burrs in the cladding shall be identified as nonconforming. Nonconforming items are segregated and evaluated within the NAC International Quality Assurance Program, and assigned one of the following dispositions: "Use-As-Is," "Rework/Repair" or "Reject." Only material that is determined to meet all applicable conditions of the license will be accepted. Embedded pieces of B4C matrix material are not considered foreign material, but such material shall be removed from the surface of the Boral. Scratches, creases or other surface indications are acceptable on the cladding of the Bora), but exposure of the core through the cladding surface of the sheet is not acceptable.

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8.1.5.3.6 Metal Matrix Composites

• Borated metal matrix composite (MMC) material can be produced by powder metallurgy, casting or thermal spray methods and consists of fine boron carbide particles in a matrix of aluminum.

Borated MMC material is a metallurgically bonded matrix, low porosity product. Borated metal matrix composites rely on a fine (average 10-40 micron) boron carbide particle size to achieve a uniform boron distribution. Specifications on the boron carbide particle size in MMCs are included in Section 8.1.5.3.11. MMCs are credited with an effectiveness of 90% of the specified minimum areal density of 10B in the borated MMC material based on acceptance and qualification testing of the material as described in Sections 8.1.5.3.8 and 8.1.5.3.10. Visual inspections of the sheets of borated MMC material will be based on Aluminum Association recommendations, as applicable-i.e., blisters and/or widespread rough surface conditions such as die chatter or porosity shall be identified as nonconforming. Nonconforming items are segregated and evaluated within the NAC International Quality Assurance Program, and assigned one of the following dispositions: "Use-As-Is," "Rework/Repair" or "Reject." Only material that is determined to meet all applicable conditions of the license will be accepted.

Local or cosmetic conditions such as scratches, nicks, die lines, inclusions, abrasion, isolated pores or discoloration are acceptable based on material neutron attenuation and thermal performance not being impacted by minor fabrication anomalies. Metal matrix composites may be encased by aluminum.

8.1.5.3.7 Borated Aluminum Borated aluminum material is a direct chill cast metallurgy product with a uniform fine dispersion of discrete boron particles in a matrix of aluminum. Borated aluminum material is a metallurgically bonded matrix, low porosity product. Borated aluminum is credited with an effectiveness of 90% of the specified minimum areal density of 10B in the borated aluminum material based on acceptance and qualification testing of the material as described in Section 8.1.5.3.8 and Section 8.1.5.3.10. Visual inspections of the sheets of borated aluminum material will be based on Aluminum Association recommendations, as applicable-i.e., blisters and/or widespread rough surface conditions such as die chatter or porosity shall be identified as nonconforming. Nonconforming items are segregated and evaluated within the NAC International Quality Assurance Program, and assigned one of the following dispositions: "Use As-Is," "Rework/Repair" or "Reject." Only material that is determined to meet all applicable conditions of the license will be accepted. Local or cosmetic conditions such as scratches, nicks, die lines, inclusions, abrasion, isolated pores or discoloration are acceptable based on material neutron attenuation and thermal performance not being impacted by minor fabrication anomalies.

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• 8.1.5.3.8 Metal Matrix and Borated Aluminum Neutron Absorber Tests

Thermal Conductivity Testing Thermal conductivity qualification testing of the neutron absorber materials shall conform to ASTM El 225, ASTM El 461, or an equivalent method. The testing shall be performed on test coupons taken from production material. Note that thermal conductivity increases slightly with temperature increases.

• Sampling will initially be one test per lot and may be reduced if the first five tests meet the specified minimum thermal conductivity. Additional tests may be performed on the material from a lot whose test result does not meet the required minimum value, but the lot will be rejected if the mean value of the tests does not meet the required minimum value.

• Upon completion of 25 tests of a single type of neutron absorber material having the same aluminum alloy matrix and boron content (in the same compound), further testing may be terminated if the mean value of all of the test results minus two standard deviations meets the specified minimum thermal conductivity. Similarly, testing may be terminated if the matrix of the material changes to an alloy with a larger coefficient of

• thermal conductivity, or if the boron compound remains the same, but the boron content is reduced.

In the Chapter 3 thermal analyses, the neutron absorber is evaluated as a 0.125-in nominal thickness sheet for the PWR fuel basket and a 0.10-in nominal thickness sheet for the BWR fuel basket.

The required minimum thermal conductivities for the MAGNASTOR neutron absorbers are as follows:

Minimum Effective Thermal Conductivity - BTU/(hr-in-°F)

Neutron Absorber Radial Axial Type 100°F 500°F 100°F 500°F Type 1 1.503 1.972 3.295 3.669 Type 2 3.12 3.21 4.31 4.65 Note: Type 1 thermal conductivity for Neutron Absorber is required for PWR or BWR basket with a maximum heat load of22 kW. Type 2 thermal conductivity is required for PWR basket with a maximum heat load of 23 kW.

The neutron absorber thermal acceptance criterion will be based on the nominal sheet thickness.

Surface anomalies increase radiation heat transfer and have insignificant influence on thermal

• conductivity, permitting acceptance of minor surface defects without additional material testing.

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Additional thermal conductivity qualification testing of neutron absorber material is not required

• if certified quality-controlled test results (from an NAC approved supplier) that meet the specified minimum thermal conductivity are available as referenced documentation.

Yield Strength Testing Yield strength qualification testing of the neutron absorber shall conform to ASTM Test Method B557/B557M, E8 or E21.

Neutron absorber material yield strength must be equal to or greater than 5 ksi at 70°F consistent with 1100-0 aluminum alloy per Table 2.2.1-14. The material yield strength at 700°F from Table 2.2.1-14 is applied as a temperature-independent value for the structural evaluation of the neutron absorber. This yield strength assures that the material will maintain its form when subjected to no1mal, off-normal and hypothetical accident condition loads.

The neutron absorber yield strength acceptance criterion will be based on the absorber meeting the specified nominal sheet thickness. Control and limitations on the neutron absorber boron content (primary driver to material structural performance) permits acceptance without additional material yield strength acceptance testing.

Additional yield strength qualification testing of neutron absorber material is not required if certified quality-controlled test results (from an NAC approved supplier) that meet the specified minimum yield strength are available as referenced documentation.

• Testing by Neutron Attenuation Acceptance testing shall be performed to ensure that neutron absorber material properties for sheets in a given production run are in compliance with the materials requirements for the MAGNASTOR fuel baskets and that the process is operating in a satisfactory manner.

Statistical tests will be run to augment findings relating to isotopic content, impurity content or uniformity of the JOB distribution.

• Determination of neutron absorber material acceptance shall be performed by neutron attenuation testing. Neutron attenuation testing of the final product, or the coupons, shall compare the results with those for calibrated standards, which may be composed of homogeneous or heterogeneous materials. The heterogeneous standard will be calibrated to a recognized standard ( e.g., homogeneous material such as ZrB2 plate material or a NIST-produced standard) or by attenuation of a thermal neutron beam correlated to the known cross-section of JOB at the beam energies. These tests shall include a statistical sample of finished product or test coupons taken from each lot of material to verify the presence, uniform distribution and the minimum areal density of JOB.

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• The 10B areal density is measured using a collimated thermal neutron beam of up to 2.54 cm in diameter, with a tolerance of 10 percent.

• Based on the MAGNASTOR required minimum effective areal density of JOB - 0.036, 0.030 or 0.027 g/cm2 for the PWR basket and 0.027, 0.0225 or 0.020 g/cm2 for the BWR basket - and the credit taken for the 10B for the criticality analyses, i.e., 90% for borated aluminum alloys and for borated metal matrix composites, a required minimum areal density for the as-manufactured neutron absorber sheets is established.

• Test locations/coupons shall be well distributed throughout the lot of material, particularly in the areas most likely to contain variances in thickness, and shall not contain unacceptable defects that could inhibit accurate physical and test measurements.

• The sampling plan shall require that each of the first 50 sheets of neutron absorber material from a lot, or a coupon taken therefrom, be tested. Thereafter, coupons shall be taken from 10 randomly selected sheets from each set of 50 sheets. This 1 in 5 sampling plan shall continue until there is a change in lot or batch of constituent materials of the sheet (i.e., boron carbide powder or aluminum powder) or a process change. A measured value less than the required minimum areal density of JOB during the reduced inspection is defined as nonconforming, along with other contiguous sheets, and mandates a return

• to 100% inspection for the next 50 sheets. The coupons are indelibly marked and recorded for identification. This identification will be used to document the neutron absorber material test results, which become part of the quality record documentation package.

• The minimum areal density specified shall be verified for each lot at the 95% probability, 95% confidence level (also expressed as 95/95 level) or better. The following illustrates one acceptable method:

- The acceptance criterion for individual plates is determined from a statistical analysis of the test results for that lot. The minimum 10B areal densities determined by neutron attenuation are converted to volume density, i.e., the neutron attenuation measurement or the maximum thickness of the coupon. The lower tolerance limit of 10B volume density is then determined - defined as the mean value of 10B volume density for the sample, less K times the standard deviation, where K is the one-sided tolerance limit factor for a normal distribution with 95% probability and 95%

confidence.

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Finally, the minimum specified value of JOB areal density is divided by the lower

• tolerance limit of JOB volume density to arrive at the minimum plate thickness that provides the specified JOB areal density.

Any plate that is thinner than this minimum or the minimum design thickness, whichever is greater, shall be treated as nonconforming, with the following exception.

Local depressions are acceptable, as long as they total no more than 0.5% of the area on any given plate and the thickness at their location is not less than 90% of the minimum design thickness.

• All neutron absorber material acceptance verification will be conducted in accordance with the NAC International Quality Assurance Program. The neutron absorber material supplier shall control manufacturing in accordance with the key process controls via a documented quality assurance system (approved by NAC or NAC's approved fabricator),

and the designer shall verify conformance by reviewing the manufacturing records.

• Nonconforming material shall be evaluated within the NAC International Quality Assurance Program and shall be assigned one of the following dispositions:

"Use-As-Is," "Rework/Repair" or "Reject." Only material that is determined to meet all applicable conditions of the license will be accepted.

• 8.1.5.3.9 Boral Neutron Absorber Tests

The Bora! neutron absorbing material is an aluminum matrix material formed from aluminum and boron-carbide. The mixing of the aluminum and boron-carbide powder forming the neutron absorber material is controlled to assure the required 10B areal density. The constituents of the neutron absorber material shall be verified by chemical testing and by dimensional measurement to ensure the quality of the finished plate or sheet. The results of all neutron absorber material tests and inspections, including the results of wet chemistry coupon testing, are documented and become part of the quality records documentation package for the fuel tube and basket assembly.

The manufacturing process of Boral consists of several steps. The initial step is the mixing of the aluminum and boron carbide powders that form the core of the finished material. The amount of each powder is a function of the desired JOB areal density. The methods used to control the weight and blend the powders are proprietary processes of the manufacturer.

After manufacturing, test samples from each Bora) batch of neutron absorber sheets shall be tested using wet chemistry techniques to verify the presence and minimum weight percent of JOB.

The tests shall be performed in accordance with approved written procedures.

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• The Bora! neutron absorber sampling plan is selected to demonstrate a 95/95 statistical confidence level in the neutron absorber sheet material in compliance with the specification. In addition to the specified sampling plan, each sheet of material is visually and dimensionally inspected using at least six measurements on each sheet. The sampling plan is supported by written and approved procedures.

The sampling plan requires that a coupon sample be taken from each of the first 100 sheets of Bora! neutron absorber material. Thereafter, coupon samples are taken from 20 randomly selected sheets from each set of 100 sheets. This 1 in 5 sampling plan continues until there is a change in lot or batch of constituent materials of the sheet (i.e., boron carbide powder, aluminum powder, or aluminum extrusion) or a process change. If either of these circumstances occurs, the sampling plan reverts back to a coupon sample being taken from each of the first 100 sheets of absorber material, followed by the 20 randomly selected sheets from each set of 100 sheets. The sheet samples are indelibly marked and recorded for identification. This identification is used to document neutron absorber test results, which become part of the quality record documentation package.

Wet Chemistry Testing Wet chemistry testing of the test coupons obtained from the sampling plan is used to verify the 10B content of the neutron absorber material. Wet chemistry testing is applied because it

• provides an accurate and practical direct measurement of the boron and B4C content of metal

materials.

An approved facility with chemical analysis capability, which could include the neutron absorber vendor's facility, shall be selected to perform the wet chemistry tests. Personnel performing the testing shall be trained and qualified in the process and in the test procedure.

Wet chemistry testing is performed by dissolving the aluminum in the matrix, including the powder and cladding, in a strong acid, leaving the B4C material. A comparison of the amount of B4C material remaining to the amount required to meet the 10B content specification is made using a mass-balance calculation based on sample size.

A statistical conclusion about the neutron absorber sheet from which the sample was taken and that batch of neutron absorber sheets may then be drawn based on the test results and the controlled manufacturing processes.

The adequacy of the wet chemistry method is based on its use to qualify the standards employed in neutron blackness testing. The neutron absorption performance of a test material is validated based on its performance compared to a standard. The material properties of the standard are NAC International 8.1-19 MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

demonstrated by wet chemistry testing. Consequently, the specified test regimen provides adequate assurance that the neutron absorber sheet thus qualified is acceptable.

The wet chemistry test results shall be considered acceptable if the 10B areal density is determined to be equal to, or greater than, that specified on the fuel tube License Drawings.

Failure of any coupon wet chemistry test shall result in 100% sampling, as described in the sampling plan, until compliance with the acceptance criteria is demonstrated.

Yield Strength Testing Yield strength qualification testing of the neutron absorber shall conform to ASTM Test Method B 557 /B 557M, E8 or E2 l. For Bora!, a laminated absorber, yield strength credited in the structural analysis was limited to the outer aluminum cover sheets. Therefore, only the cover sheet must be shown to meet the required strength.

8.1.5.3.10 Qualification Testing of Metal Matrix and Borated Aluminum Neutron Absorber Material Qualification tests for each MAGNASTOR System neutron absorber material and its set of manufacturing processes shall be performed at least once to demonstrate acceptability and durability based on the critical design characteristics, previously defined in this section.

The licensed service life will include a range of environmental conditions associated with short

• term transfer operations, normal storage conditions, as well as off-normal and accident storage events. Additional qualification testing is not required for a neutron absorber material previously qualified, i.e., reference can be provided to prior testing with the same, or similar, materials for similar design functions and service conditions.

• Qualification testing is required for: ( 1) neutron absorber material specifications not previously qualified; (2) neutron absorber material specifications previously qualified, but manufactured by a new supplier; and (3) neutron absorber material specifications previously qualified, but with changes in key process controls. Key process controls for producing the neutron absorber material used for qualification testing shall be the same as those to be used for commercial production.

• Qualification testing shall demonstrate consistency between lots (2 minimum).

• Environmental conditions qualification will be verified by direct testing or by validation by data on the same, or similar, material, i.e., the neutron absorber material is shown to not undergo physical changes that would preclude the performance of its design functions. Conditions encountered by the neutron absorber material may include: short term immersion in water, exposure to chemical, temperature, pressure, and gamma and neutron radiation environments. Suppliers' testing will document the durability of

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• neutron absorber materials that may be used in the MAGNASTOR System by demonstrating that the neutron absorber materials will not incur significant damage due to the pressure, temperature, radiation, or corrosion environments or the short-term water immersion that may occur in the loading and storage of spent fuel.

• Thermal conductivity and yield strength qualification testing shall be as previously described in Section 8.1.5.3.8.

• The uniformity of the boron carbide distribution in the material shall be verified by neutron attenuation testing of a statistically significant number of measurements of the areal density at locations distributed throughout the test material production run, i.e., at a minimum from the ends and the middle of the run. The sampling plan must be designed to demonstrate 95/95 compliance with the absorber content requirements. Details on acceptable neutron attenuation testing are previously provided in this section for Acceptance Testing. Alternate test methods may be employed provided they are validated (benchmarked) to neutron attenuation tests.

• One standard deviation of the neutron attenuation test sampling results shall be less than 10% of the sample mean. This requirement provides additional assurance that a consistent product is achieved by the manufacturing process.

• A material qualification report verifying that all design requirements are satisfied shall be

• prepared.

• Key manufacturing process controls in the form of a complete specification for materials and process controls shall be developed for the neutron absorber material by the supplier and approved by NAC to ensure that the product delivered for use is consistent with the qualified material in all respects that are important to the material's design function.

• Major changes in key manufacturing processes for neutron absorber material shall be controlled by mutually agreed-upon process controls established by the certificate holder/purchaser and the neutron absorber supplier. These process controls will ensure that the neutron absorber delivered will always be consistent with the qualification test material in any and all respects that are important to the neutron absorber's safety characteristics. Changes in the agreed-upon process controls may require requalification of those parts of the qualification that could be affected by the process changes. Typical changes covered by the agreed-upon process controls include:

- Changes that could adversely affect mechanical properties ( e.g., change in thermal conductivity, porosity, material strength, change of matrix alloy, boron carbide content, increase in the B4C content above that used in previously qualified material, etc.);

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- Changes that could affect the uniformity of boron (e.g., change to mixing process for

• aluminum and boron carbide powders, change in stirring of melt, change in boron precipitate phase, etc.).

• Minor neutron absorber material processing changes, i.e., roller machine hardware or final sheet cutting methods, water jet, shear cut, etc., may be determined to be acceptable on the basis of engineering review without additional qualification testing, if such changes do not adversely affect the particle bonding microstructure, i.e., the durability or the uniformity of the boron carbide particle distribution, which is the neutron absorber effectiveness.

• Nonconforming material shall be evaluated within the NAC International Quality Assurance Program and shall be assigned one of the following dispositions:

"Use-As-Is," "Rework/Repair" or "Reject." Only material that is determined to meet all applicable conditions of the license will be accepted.

8.1.5.3.11 Additional Material Specifications Boron carbide particles for MM Cs shall have an average size in the range of 10-40 microns and no more than 10% of the particles shall be over 60 microns. The material shall have negligible interconnected porosity exposed at the surface or edges.

Open porosity for borated aluminum and borated MMC neutron absorber material must be no

• greater than 0.5% unless qualification tests are performed to ensure that blisters are not produced under submerging and subsequent vacuum drying conditions.

Chemical composition of the boron carbide powder must meet the requirements of Table 1 of ASTM C 750-03, Type 3. Additional chemical requirements, applicable to a particular absorber material, may be placed on the boron carbide powder as a result of the "key manufacturing process controls" invoked by Section 8.1.5.3.10. Additional requirements may include, but are not limited to, upper limits on fluorine and chlorine content.

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• Table 8.1-1 Neutron Absorber Material Minimum 10B Loading

Required Minimum Effective Required Minimum Actual Neutron Areal Density % Credit Used in Areal Density Absorber (108 ~ /cm2) Criticality (108 ~ /cm2)

Type PWR Fuel 8WR Fuel Analyses PWR Fuel 8WR Fuel Borated 0.036 0.027 0.04 0.03 Aluminum Alloy 0.030 0.0225 90 0.0334 0.025 0.027 0.020 0.03 0.0223 Borated MMC 0.036 0.027 0.04 0.03 0.030 0.0225 90 0.0334 0.025 0.027 0.020 0.03 0.0223 Baral 0.036 0.027 0.048 0.036 0.030 0.0225 75 0.04 0.030 0.027 0.020 0.036 0.0267

Table 8.1-2 Mechanical Properties of Neutron Absorber

• Property (units) Values at Temperature (°F) 70 Ultimate Tensile Strength, Su (ksi) a 13.1 Yield Strength, Sy (ksi) a 5.0

a Equal to aluminum alloy 1100-0 properties

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8.1.6 Shielding Tests

• Chemical-copper grade lead (ASTM B29) poured in the annulus between the inner and outer stainless steel shells of the MAGNA TRAN transpo11 cask body provide the primary radial gamma shielding. Thirty (30) individual Type 304 stainless steel assemblies are used to enclose neutron shield material and these assemblies are positioned around the circumference and along the length of the cask body to provide neutron shielding for the cask. Testing of the gamma shielding material for effectiveness and confirming the as-poured effectiveness of the neutron shield material is discussed in the following paragraphs.

8.1.6.1 Gamma Shielding Test A gamma scan test of the stainless steel and lead radial shielding composing the MAGNA TRAN transport cask body will be performed prior to installation of the copper and aluminum cooling fins, the neutron shield weld studs, and the stainless steel enclosed neutron shield assemblies to verify the installed effectiveness of the cask's gamma shielding. The test will be performed in accordance with approved written procedures. The gamma scan test involves using a detector and a 6°Co source to continuously scan, or probe, over the entire cask surface along the length of the cask cavity. The 6°Co source strength will be sufficiently intense to produce a count rate that equals or exceeds three times the background count rate on the external surfaces of the cask. The maximum scan path spacing will be 2.5 inches and the scanning speed will be 4.5 feet/minute, or

• less. All scanning will be on a 2-in. grid pattern (when using a 3-in. scintillation type detector),

and the specified count time ( dwell time) will be greater than one minute, based on the measured response time with the mockup. The acceptance criterion for the gamma scan test will be that the shielding effectiveness of the cask body is greater than the shielding effectiveness of a lead and stainless steel shielding mockup. The shielding mockup will include the stainless steel and lead thicknesses that are equivalent to the minimum thickness specified for each material on the License Drawings. The shielding mockup will be produced by using the approved fabrication techniques for the cask. The acceptance criteria for the gamma shielding test is represented by the following formula:

where:

Cs= counts per second at the testing area of the cask, and

Cr= counts per second of the mockup coupon.

In any area which produces a measured count rate that exceeds that established by the mockup coupon shall be rejectable, the cask body will be rejected. The areas or components of the cask NAC International 8.1-24 MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• body that are rejected will be evaluated to determine the corrective action to be taken. Any repaired areas and components, including areas surrounding the repaired area that might have been affected by the repair, will be retested to the original acceptance criteria prior to final acceptance.

8.1.6.2 Neutron Shielding Test The MAGNA TRAN neutron shield is provided by an assembly of 30 individual neutron shield sections encased in welded stainless steel enclosures. The mechanical and thermal properties of NS-4-FR neutron shielding material are provided in Chapters 2 and 3, respectively. Dimensional inspection of each of the individual neutron shield assemblies containing the NS-4-FR shall be performed to verify that the required minimum thickness specified on the License Drawings is met prior to final closure of the component. Each mixed batch ofNS-4-FR shall be tested to verify that the material composition (aluminum and hydrogen), boron concentration and neutron shield density meet the requirements specified in Chapters 2 and 3 and on the License Drawings.

Material testing shall be performed by qualified laboratories in accordance with written and approved procedures. Material composition, boron concentration and density data for each lot of neutron shield material shall become part of the quality record documentation package. Samples of each lot of neutron shield material shall be maintained as part of the quality record documentation package.

• The installation of the NS-4-FR into the thirty individual neutron shield subassemblies shall be

performed in accordance with written, approved and qualified procedures. The installation procedures shall ensure that mix ratios and mixing methods are controlled in order to achieve proper material composition, boron concentration and distribution, and that pouring techniques are controlled in order to prevent gaps, or unacceptable voids, from occurring in the NS-4-FR.

The neutron shield thickness will be readily verifiable prior to final stainless steel enclosure plate installation. Installation procedures shall be qualified by the use of mockups.

8.1.6.3 Neutron and Gamma Shield Effectiveness Test Following first fuel loading and prior to transport, a neutron and gamma shielding effectiveness test is performed for each fabricated MAGNA TRAN cask in accordance with approved written procedures. The purpose of the test is to document the effectiveness of the neutron and gamma shielding materials. For this test, the cask is loaded with a canister containing spent nuclear fuel that is welded, drained, vacuum dried, and backfilled with helium.

Calibrated neutron and gamma dose rate meters shall be used to measure the neutron and gamma dose rate at contact with the outer shell of the neutron shield and at 2.3 m from the surface

• (equivalent to 2 m from the sides of the railcar). Dose measurement points are established on the

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external surface of the shell at 30° intervals and at five points along the height of the shield (a

• total of 60 measuring points). In addition, neutron and gamma dose rate measurements are made at the trunnion areas above the neutron shield, at four points below the neutron shield, and at the edges and center of the cask top and bottom surfaces. Dose rates at the top and bottom of the cask are measured with the impact limiters installed. The dose rates measured at contact and at 2.3 m are recorded on the test data sheet. Additional data recorded for the shielding effectiveness test include the total power of the loaded fuel assemblies; date, time and location of test; identification and calibration of instrumentation; and identification of test engineer and operators. To enable the measured dose rates to be evaluated, the burnup and cool time for the actual fuel assemblies loaded into the cask are determined and recorded. From this fuel history data, the total actual neutron and gamma source terms are estimated by using ORI GEN or similar calculations. Neutron and gamma source terms applied in these dose rate calculations are to be based on the fuel type, MTU, burnup, and initial enrichment of the fuel loaded in the cask at the time dose rate measurements are taken. The package configuration (including materials and geometry) to be applied in the dose rate calculation must represent the minimum shielding effectiveness configuration discussed in Chapter 5. Utilizing the neutron and gamma source terms estimated (i.e., calculated) gamma and neutron dose rates are developed in order to properly evaluate the measured dose rates.

If the measured dose rates exceed the estimated dose rates corresponding to the MAGNATRAN

• fuel contents, the cask User shall notify the NRC in accordance with 10 CFR 71.95. Appropriate corrective measures shall be taken including unloading the TSC from the MAGNA TRAN and correction of identified shielding deficiencies. Following corrective actions, the shield effectiveness test shall be re-performed to the original acceptance criteria prior to final cask acceptance for transport operations.

8.1. 7 Thermal Acceptance Test

Prior to acceptance of each MAGNA TRAN transport cask, a thermal test using electric heaters will be performed on a fabricated packaging to verify that the fabricated and assembled transport cask possesses the heat rejection capabilities evaluated in the thermal analyses in Chapter 3. The thermal test will be performed in accordance with approved written procedures.

8.1.7.1 Thermal Test Setup

A typical thermal test set-up is shown in Figure 8.1-1. As depicted, the thermal test will be performed with the cask positioned horizontally on a test frame. The transport impact limiter or equivalent insulating material will be installed on each end of the cask to simulate the transport configuration. The cask will be located in a covered building in a still environment. The NAC International 8.1-26 MAGNATRAN Transport Cask SAR August 2021 Docket No. 71-9356 Revision 21 B

• contents heat will be simulated by using electric heaters to apply 23 kW to the inner surface of the transport cask. The electric heating elements will be mounted on a metallic shell and will extend for the full length of the cask cavity except for approximately 6 inches from each end.

The metallic heater-mounted test shell will radiate and conduct heat to the inner surface of the cask cavity. The cask bottom or top will receive negligible heat. Due to the small radial gap between the metallic heater-mounted test shell and the inner cask surface, the heat transfer by convection is negligible compared to the heat transfer by radiation and conduction. Electrical heaters are spaced to provide a uniform heat flux, circumferentially to the cask wall and permit power input to be measured for heating element.

As described below and Figure 8.1-1, six calibrated thermocouples will be installed on the inner surface of the MAGNA TRAN transport cask and six calibrated thermocouples will be installed on the cask external surface at the same angular and axial locations as the thermocouples on the cask inner surface.

• Four thermocouples located at approximately 90° intervals (see Figure 8.1-1) on the inner shell surface 88 inches from the inner surface of the cask bottom

• One thermocouple at the upper region (180° location as shown in Figure 8.1-1) of the inner shell surface at 21 inches from the inner surface of the cask bottom

• One thermocouple at the upper region (180° location as shown in Figure 8.1-1) of the inner shell surface at 154 inches from the inner surface of the cask bottom

• Four thermocouples located at approximately 90° intervals (see Figure 8.1-1) on the neutron shell outer surface with the axial location at 88 inches from inner surface of cask bottom and at the thermocouples on the inner shell surface

• One thermocouple at the upper region of the neutron shield shell (180° location as shown in Figure 8.1-1) with the axial location at 154 inches from the inner surface of cask bottom and at the same angular location as the thermocouple on the inner shell surface

• One thermocouple at the upper region of neutron shield shell surface (180° location as shown in Figure 8.1-1) with the axial location at 154 inches from inner surface of cask bottom and at the same angular location as the thermocouple on the inner shell surface

An additional thermocouple is required to measure the ambient temperature. The thermocouples will be attached to strip chart recorders, which may have multiple input channels, or another similar device to allow for continuous monitoring and recording of temperatures during the test.

These records will be part of the quality assurance acceptance records for the cask under test.

The nominal test conditions are ambient temperature and with the initial cask body temperatures being 70°F, no solar insolation, still air, and no external radiant heat sources. At these test conditions, the cask surface temperature will be bounded by the calculated steady-state

• equilibrium temperature of l 78°F. It will take about 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> for the system to approach steady

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state condition. The thermal test procedure will provide a thermal transient heat-up curve to

• show the time at which equilibrium is expected to be established, and a table, or set of curves, that correlates equilibrium neutron shield assembly surface temperature with a range of ambient temperatures. For purposes of the thermal test, equilibrium temperature is assumed to be established when the change in neutron shield assembly surface temperature no longer exceeds 2°F in a two-hour period.

8.1.7.2 Thermal Test Acceptance Criteria The purpose of the thermal test is to confirm that the heat rejection capabilities of the as-built MAGNATRAN transport cask are acceptable. Cask thermal test acceptance is based on the demonstration that the measured temperature gradients are less than, or equal to, the thermal gradients calculated in the thermal analyses, as described in Section 8.1.7.3, and that the total heat rejection rate is equal to, or greater than, the cask design basis heat rejection rate.

8.1.7.3 Thermal Analyses for Thermal Test A three dimensional half-symmetry model based on the full-length model described in Section 3.4.1.1.1 is utilized to perform steady state and transient analyses for the usage of the thermal test. The steady state analysis evaluates the temperatures at the thermocouple locations of the thermal test, while the transient analysis identifies the duration for the system to reach a steady state condition.

• The model boundary conditions described in Section 3.4.1.1.1 were modified to reflect the actual testing conditions of the transport cask. The main boundary conditions were listed below:

1) A uniform heat flux is applied to the entire inner surface of the cask inner shell based on the design heat load of 23kW.

2) Convection film coefficient for the model, as described in Section 3.2.3, is used.

3) No solar insolance is considered in this analysis.

4) An ambient temperature of 70°F is considered.

In order to compare the analytical results with the thermal test results, the maximum component temperatures and temperatures at three axial locations of the cask inner and outer surfaces are post processed for the steady state condition. The transient thermal analysis confirms that approximately 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> are needed for the cask to reach thermal steady state condition.

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• Figure 8.1-1 Thermal Test Arrangement

270° go* i REF REF 276' :=-+-=--= i\\

REF i\\

i ~- *12° REF o* REF REF LOCATION AT 21" AND 154" LOCATION AT 88"

rAMBIENT

• r--154" REF

• I 188" REF-~

CASK TOP I-21" REF

CASK BOTTOM

---

*--------------I

1 t:::j

THERMOCOUPLE LOCATION

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8.1.8 Miscellaneous Tests The MAGNA TRAN transport cask has several sealed and enclosed components that require exclusion of foreign materials, including water, to prevent deterioration of the materials contained therein. These components include the thirty individual neutron shield sections filled with NS-4-FR neutron shielding material enclosed in stainless steel shells, and the transport impact limiters which are filled with energy absorbing redwood and balsa wood inside of a stainless steel shell. Following final closure of the 30 individual neutron shielding sections, and the front and rear impact limiters in their welded stainless steel shells, a dye penetrant examination of all of the closure welds shall be performed to verify that there is no potential for leakage of foreign material including rain water. Note that the MAGNATRAN is designed to be loaded and unloaded dry without immersion in a spent fuel pool, and therefore, the neutron shield subassemblies will not be exposed to spent fuel pool hydrostatic pressure loads.

The acceptance criteria for the dye penetrant examination shall be ASME Section III, Section NF, Article NF-5350.

As described in Section 8.2, the dye penetrant examinations shall be re-performed periodically as part of the maintenance program.

8.1.9 Packaging Identification

• Upon successful completion of the MAGNATRAN transport cask inspection and acceptance test program, each MAGNA TRAN packaging shall have a regulatory plate containing the information required per 10 CFR 71.85( c) attached to the cask body in accordance with the license drawings. A serial number shall also be included on the regulatory plate to uniquely identify the specific cask and to establish traceability to the applicable fabrication records.

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