ML19226A351

Enclosure 1: CoC No. 9341, Revision No. 8 (Letter to P.Noss Revision No. 7 of Certificate of Compliance No. 9341 for the Model No. Brr Package)

ML19226A351
Person / Time
Site:	07109341
Issue date:	08/15/2019
From:	John Mckirgan Spent Fuel Licensing Branch
To:	AREVA Federal Services
Devaser N
Shared Package
ML19226A348	List: ML19226A350 ML19226A351 ML19226A352
References
EPID L-2019-LLA-0136
Download: ML19226A351 (12)

Text

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 9341 8 71-9341 USA/9341/B(U)F-96 1 OF 12

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 Orano Federal Services LLC AREVA Federal Services LLC application dated 505 S. 336th Street, Suite 400 June 13, 2019.

Federal Way, WA 98003

4. CONDITIONS This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below.

5.

(a) Packaging (1) Model No.: BEA Research Reactor (BRR) Package (2) Description The purpose of the Model No. BRR package is to transport irradiated fuel elements or loose plates of a square fuel element from various test and research reactors. The package is comprised of a lead-shielded package body, payload basket, square loose plate box, an upper shield plug, a closure lid, upper and lower impact limiters, and utilizes American Society for Testing and Materials (ASTM) Type 304 stainless steel as its primary structural material. The package is a right circular cylinder with a dimension of 77.1 inches in length and 38 inches in diameter, not including the impact limiter attachments and the thermal shield. Lead shielding is located between two circular shells, in the lower end structure, and in the shield plug. The payload cavity has a diameter of 16 inches and a length of 54 inches.

Impact Limiters. Impact limiters are attached to each end of the package body. Each impact limiter is 78 inches in diameter and 34.6 inches in length, with a 15-inches long conical section towards the outer end. The impact limiter design consists of ASTM Type 304 stainless steel shells and polyurethane foam with an approximate density of 9 pounds per cubic foot (lb/ft3).

Fuel Baskets. There are six baskets used with the package, one for each type of fuel transported and one for isotope production targets. The baskets are made from welded construction using ASTM Type 304 stainless steel in plate, bar, pipe, and tubular forms.

Each basket has a diameter of 15.63 inches and a length of 53.45 inches, and features a number of cavities that fit the size and shape of the fuel. The basket for square fuel

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 9341 8 71-9341 USA/9341/B(U)F-96 2 OF 12 5.(a) Packaging - Description (continued) accommodates two types of fuel assembly: (1) flat type fuels and (2) a 5x5 array of fuel rods enclosed within a rectangular can.

Personnel Barrier. When transporting isotope production targets, a personnel barrier is used to limit access to the package body such that personnel are prevented from touching the cask surface where the surface temperature may exceed the allowable limit for exclusive use shipments. The barrier does not have a radiological purpose.

Spacer Pedestals. For fuel elements or assemblies shorter than the length of a basket cavity, spacer pedestals are used in each cavity, as required, to support the fuel elements at the top of the basket. All spacer pedestals are made of stainless steel Square Box or Loose Plate Box. A square box accommodates square fuel loose plates. A loose plate box is used to transport up to 31 loose plates per box. The square fuel basket and loose plate box are made of stainless steel.

The package is designed to be transported as one package per conveyance, with its longitudinal axis vertical, by highway truck or by rail in exclusive use. When loaded and prepared for transport, the package is 119.5 inches in length, 78 inches in diameter (over the impact limiters), and weighs 32,000 pounds (lb).

(3) Drawings The packaging is constructed in accordance with AREVA Federal Services LLC drawings:

- 1910-01-01-SAR, BRR Package Assembly SAR Drawing, Sheets 1-5, Rev. 8

- 1910-01-02-SAR, BRR Package Impact Limiter SAR Drawing, Sheets 1-2, Rev. 1

- 1910-01-03-SAR, BRR Package Fuel Baskets SAR Drawing, Sheets 1-4, Rev. 6

- 1910-01-04-SAR, BRR Package Isotope Target Basket SAR Drawing, Sheets 1-2, Rev. 1 (b) Contents (1) Type and form of material Irradiated MURR Fuel Element. Irradiated University of Missouri Research Reactor (MURR) fuel element to a maximum burnup of 180 megawatt-day (MWD) or a depletion of 30.9% of Uranium-235 (235U). The minimum cooling time is 180 days after reactor shutdown. Each MURR element contains 24 fuel plates. Each fresh MURR fuel element contains 775.0 +/- 7.8 g 235U. The enrichment range is 93 +/-1 wt.% 235U. The MURR element overall length, including irradiation growth, is 32.75 inches. The maximum decay heat per fuel element is 158 watts (W). The maximum number of fuel elements per basket is 8. The bounding weight of one element is 15 lb. Table 1.1 includes characteristics of a pre-irradiated MURR fuel element.

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 9341 8 71-9341 USA/9341/B(U)F-96 3 OF 12 5.(b)(1) (i) Type and form of material - Irradiated MURR Fuel Element (continued)

Table 1.1. MURR - Key Fuel Element Parameters Maximum active fuel length (inches) 24.8 Overall length (inches) 32.75 Minimum cladding thickness (inch) 0.008 Nominal fuel matrix thickness (inch) 0.02 Fuel matrix UAlx Cladding material Aluminum Maximum 235U per element (g) 782.8 Maximum enrichment (wt.%) 94.0 Maximum U per fuel plate (g) 235 46.0 (ii) Irradiated MITR-II Fuel Element. Irradiated Massachusetts Institute of Technology Research Reactor (MITR-II) fuel element to a maximum burnup of 165 MWD or a 235Udepletion of 43.9%. The minimum cooling time is 120 days after reactor shutdown. Each MITR-II element contains 15 fuel plates. Each fresh MITR-II element contains 510.0 +3.0/-10.0 g 235U, which is 500 - 513 g 235U. The enrichment range is 93 +/-1 wt.% 235U. The MITR-II element overall length, including irradiation growth, is 26.52 inches. The maximum decay heat per element is 150 W. The maximum number of fuel elements per basket is 8. The bounding weight of one element is 10 lb. Table 1.2 includes the key parameters for a pre-irradiated MITR-II fuel element.

Table 1.2. MITR-II - Key Fuel Element Parameters Maximum active fuel length (inches) 22.76 Overall length (inches) 26.52 Minimum cladding thickness (inch) 0.008 Nominal fuel matrix thickness (inch) 0.03 Maximum fuel matrix width (inches) 2.171 Fuel matrix UAlx Cladding material Aluminum Maximum 235U per element (g) 513 Maximum enrichment (wt.%) 94.0 Maximum 235U per fuel plate (g) 34.3 (iii) Irradiated ATR Fuel Element. Irradiated Advanced Test Reactor (ATR) fuel element to a maximum burnup of 480 MWD or a 235U depletion of 58.6%. The minimum cooling time is 1,670 days (4.6 years) after reactor shutdown. Each ATR fuel element contains 19 plates. The YA fuel element has 19 plates, but only 18 contain fuel.

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 9341 8 71-9341 USA/9341/B(U)F-96 4 OF 12 5.(b)(1) (iii) Type and form of material - Irradiated ATR Fuel Element (continued)

There are two general classes of ATR fuel element, XA and YA. The enrichment range is 93 +/- 1 wt.% 235U. The XA fuel element has a fresh fuel loading of 1,075 +/- 10 g 235U. The YA fuel element has a fresh fuel loading of 1,022.4 +/- 10 g 235U. A second YA fuel element design (YA-M) has the side plate width reduced by 15 mils. The ATR element overall maximum length, after removal of the end box structures, 51.0 inches. The maximum number of fuel elements per basket is 8. The bounding weight of one element is 25 lb. The maximum decay heat per element is 30 W. Table 1.3 includes characteristics of a pre-irradiated ATR fuel element.

Table 1.3. ATR - Key Fuel Element Parameters Maximum active fuel length (inches) 48.77 Overall length (inches) 51 Minimum cladding thickness for Plate 1 (inch) 0.018 Minimum cladding thickness for Plates 2-18 (inch) 0.008 Minimum cladding thickness for Plate 19 (inch) 0.018 Nominal fuel matrix thickness (inch) 0.02 Fuel matrix UAlx Cladding material Aluminum Maximum 235U per element (g) 1,085 Maximum enrichment (wt.%) 94.0 Maximum 235U per fuel plate (g) 85.2 (iv) Irradiated TRIGA fuel elements. Table 1.4 includes the dimensions of pre-irradiated Training, Research, Isotopes, General Atomics (TRIGA) fuel elements. The TRIGA fuel matrix is uranium mixed with zirconium hydride. The BRR package is limited to the transportation of the following types of TRIGA fuel:

1. Standard 100 series.

2. Instrumented 200 series. The fuel region is as the same as 100 series but contain thermocouples used to measure temperature during reactor operation.

Instrumented rods may be longer than 100 series.

3. Fueled Follower Control Rods (FFCR) (300 series). The rods contain boron carbide neutron absorber outside the active fuel region.

4. Cluster Rods (400 series). It is typically built with three or four cluster rods to make a cluster assembly.

5. Instrumented Cluster Rods (500 series). Fuel is the same as cluster rod but thermocouples used to measure temperature during reactor operation.

Instrumented cluster rods may be longer.

Cluster rods (i.e., TRIGA fuel series 400 and 500) must be disassembled from the cluster assembly for transport in the BRR package.

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 9341 8 71-9341 USA/9341/B(U)F-96 5 OF 12 5.(b)(1) (iv) Type and form of material - Irradiated TRIGA fuel elements (continued)

Table 1.4. Characteristics of Pre-Irradiated TRIGA Fuel Fuel 235 235 Fuel Rod Cladding Overall U U U U Erbium Type ID1 Cladding Length OD2 OD Thickness H/Zr Length3 (wt. % Fuel) (wt. %) (g) (g) (wt. %)

(in.) (in.) (in.) (in.) (in.)

101 14 8.0 20 166 32 1.41 1.48 0.03 1.0 28.62 0 Aluminum 101 15 8.5 20 189 37 1.41 1.48 0.03 1.6 28.62 0 Standard 100 series 103 15 8.5 20 197 39 1.44 1.48 0.02 1.6 29.15 0 105 15 12 20 285 56 1.44 1.48 0.02 1.6 29.15 0 107 Stainless 15 12 20 271 53 1.4 1.48 0.02 1.6 30.14 0 109 Steel 15 8.5 70 194 136 1.44 1.48 0.02 1.6 29.15 1.2 117 15 20 20 503 99 1.44 1.48 0.02 1.6 29.93 0.5 119 15 30 20 825 163 1.44 1.48 0.02 1.6 29.93 0.9 201 Aluminum 15 8.5 20 189 37 1.41 1.48 0.03 1.6 28.78 0 Instrumented 200 203 15 8.5 20 197 39 1.44 1.48 0.02 1.6 45.5 0 205 15 12 20 285 56 1.44 1.48 0.02 1.6 45.5 0 Stainless series 207 15 12 20 271 53 1.4 1.48 0.02 1.6 45.5 0 Steel 217 15 20 20 503 99 1.44 1.48 0.02 1.6 40.35 0.5 219 15 30 20 825 163 1.44 1.48 0.02 1.6 40.35 0.9 303 15 8.5 20 163 32 1.31 1.35 0.02 1.6 44 0 Fueled Follower 305 15 12 20 237 47 1.31 1.35 0.02 1.6 44 0 Control Rods Stainless Steel 317 15 20 20 418 82 1.31 1.35 0.02 1.6 44 0.5 (FFCR) (300 series) 319 15 30 20 685 135 1.31 1.35 0.02 1.6 44 0.9 403 15 8.5 20 166 33 1.37 1.41 0.02 1.6 30.38 0 Cluster rods 405 Stainless 15 12 20 243 48 1.37 1.41 0.02 1.6 30.38 0 Steel (400 series) 417 15 20 20 427 85 1.37 1.41 0.02 1.6 30.38 0.5 419 15 30 20 710 141 1.37 1.41 0.02 1.6 30.38 0.9 503 15 8.5 20 166 33 1.34 1.41 0.02 1.6 45.5 0 Instrumented 505 15 12 20 243 48 1.34 1.41 0.02 1.6 45.5 0 Stainless cluster rods 517 Steel 15 20 20 427 85 1.34 1.41 0.02 1.6 45.5 0.5 (500 series) 519 15 30 20 710 141 1.34 1.41 0.02 1.6 45.5 0.9 1

General Atomics catalog numbers are not necessarily unique. TRIGA elements with the same ID could have different fuel parameters. Table 1.4 includes two variants of the Type 101 element.

2 Outer Diameter.

3 Overall length includes 0.25 inches for irradiation growth.

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 9341 8 71-9341 USA/9341/B(U)F-96 6 OF 12 5.(b)(1) (iv) Type and form of material - Irradiated TRIGA fuel elements (continued)

The maximum length of a TRIGA fuel element, including irradiation growth, is 45.50 inches. For all fuel elements, stainless steel spacers are utilized within the TRIGA baskets. The bounding weight of any TRIGA fuel element is 10 lb. The maximum decay heat per element is 20 W. The number of TRIGA rods per element is 1. Table 1.5 includes parameters for irradiated TRIGA fuel.

Table 1.5. Maximum Burnup and Minimum Cooling Time for TRIGA Fuel Elements4 TRIGA Fuel Type (Enrichment) Maximum Burnup (MWD) Minimum Cooling Time (days) 101 (8.0%) 23 90 201/101 (8.5%) 26 90 88 350 70 250 109 52 170 34 90 203/103 27 90 39 120 205/105 33 90 38 120 207/107 33 90 71 280 217/117 52 180 34 90 122 600 91 370 219/119 63 220 34 90 303 22 90 305 32 90 58 210 317 46 150 34 90 97 420 76 290 319 55 180 34 90 503/403 23 90 505/405 33 90 60 220 517/417 47 150 34 90 101 430 79 290 519/419 56 180 34 90 4

Based on an in-core residence time of 4 years resulting on a decay heat less than or equal to 20 W. Not applicable to fuel with an in-core residence time less than 4 years with a decay heat greater than 20 W.

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 9341 8 71-9341 USA/9341/B(U)F-96 7 OF 12 5.(b)(1) Type and form of material (continued)

(v) PULSTAR Fuel. Table 1.6 includes the characteristics of the PULSTAR fuel. A 5x5 array of fuel rods enclosed within a rectangular can. Each fuel rod contains cylindrical uranium oxide fuel pellets. The weight of a PULSTAR element is 48 lb, including a spacer pedestal. The maximum heat load of the square fuel basket per compartment is 30 W.

Table 1.6. Characteristics of PULSTAR Fuel Parameter Value Nominal 235U Enrichment (%) 4.0/6.0 Fuel matrix UO2 Maximum burnup (MWD/MTU) 20,000 Decay time (years) 1.5 Maximum fuel pellet diameter (in.) 0.423 Minimum cladding thickness (in.) 0.0185 Cladding material Zirconium alloy Maximum cladding OD (in.) 0.474 Maximum active fuel length (in.) 24.1 Fuel rod pitch X (in.) 0.607 Fuel rod pitch Y (in.) 0.525 Box outer dimensions (in.) 3.15 x 2.74 Box thickness (in.) 0.06 Box material Zirconium alloy Maximum overall length (in.) 38.23 Note: Maximum length includes 0.25 in. for irradiation growth.

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 9341 8 71-9341 USA/9341/B(U)F-96 8 OF 12 5.(b)(1) Type and form of material (continued)

(vi) Square Fuel and Loose Plates (excluding PULSTAR). Table 1.7 includes the main characteristics of square fuel and square-loose-plate fuel. These types of fuel have a square, or nearly square-rectangular cross section. The flat-type fuels consist of either a uranium-oxide dispersion or uranium-silicide dispersion meat in an aluminum matrix, bonded with an aluminum alloy cladding. The maximum heat load of the square fuel basket per compartment is 30 W.

Table 1.7. Square Plate Fuel Characteristics Parameter RINSC Ohio State Miss. S&T U-Florida Purdue U-Mass (Al) U-Mass (Si) 235U 275+/-7.7 200+/-5.6 225+/-6.3 175+/-4.9 129.92+/-2.52 167+/-3.3 200+/-5.6 loading (g)

Nominal 235U 19.75 19.75 19.75 19.75 19.75 19.75 19.75 enrichment (%)

Fuel matrix U3Si2+Al U3Si2+Al U3Si2+Al U3Si2+Al U3Si2+Al UAlx U3Si2+Al Maximum burnup per fuel element 52.5 64.0 74.0 87.0 0.57 9.7 9.7 (MWD)

Minimum decay 120 120 365 120 120 1,000 1,000 time (D)

Nominal fuel meat 2.395 2.395 2.395 2.395 2.395 2.320 2.395 width (in.)

Nominal fuel meat 0.02 0.02 0.02 0.02 0.02 0.03 0.02 thickness (in.)

Nominal fuel plate 0.05 0.05 0.05 0.05 0.05 0.06 0.05 thickness (in.)

Nominal active fuel 23.25 23.25 23.25 23.25 23.25 23.25 23.25 length (in.)

Number of fuel 22 16 18 14 14 18 16 plates Maximum channel 0.099 0.127 0.139 0.117 0.175 0.119 0.122 spacing (in.)

Weight (lb) 14 12 14 10 10 12 12 Maximum overall 39.75 35.25 34.50 27.38 32.49 39.75 39.75 length (in.)

Maximum cross 3.097x3.097 3.05x3.05 3.036x3.212 2.9x2.424 3.011x3.011 3.097x3.097 3.097x3.097 section (in.)

Loose plate no no no yes yes yes no Notes:

1. U-Mass (Al) loose plates have a 235U loading of 9.28 +/- 0.18g and dimensions of 2.78 inches wide by 24.88 inches long.

2. U-Florida loose plates have a 235U loading of 12.5 +/- 0.35g and dimensions of 2.85 inches wide by 25.88 inches long.

3. Purdue loose plates have a 235U loading of 9.28 +/- 0.18g and dimensions of 2.85 inches wide by 25.88 inches long.

4. Maximum length includes 0.25 inches for irradiation growth.

5. Loose plates shall be extracted from fuel elements that meet the per-element burnup limits provided in this table.

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 9341 8 71-9341 USA/9341/B(U)F-96 9 OF 12 5.(b)(1) Type and form of material (continued)

(vii) Isotope Production Targets. Targets are irradiated in nuclear reactors to produce Co-60 and may be made of aluminum and contain a large quantity of small pellets, or they may consist of a cylindrical rod of cobalt material inside a stainless steel tube.

All targets must be placed into target holders prior to loading into the basket. There are two different payload types:

1. Payload Type 1. Type 1 consists primarily of higher-activity targets of a newer design, which may also include lower-activity targets as described under Payload Type 2. The pellets are arranged in several stacks in an annular configuration within the target body. Payload Type 1 consists of up to 10 targets, which must be loaded in the inner row of basket holes, and be arranged using a loading plan into five zones of two holes each. The maximum activity in any zone is 22,000 Ci.

A loading collar must be installed to block access to the outer row of holes before loading payload Type 1 targets. Table 1.8 includes the characteristics of payload type 1 of the isotope production targets.

Table 1.8. Characteristics of Isotope Production Targets, Payload Type 1 Parameter Value Target Diameter 1/2 inches Target Length 16 inches Cladding Material 6061-T6 aluminum alloy Target Contents 6,000 pellets (approximately)

Pellet Size 1mm diameter x 1mm thick Maximum Activity up to 14,100 Ci, Co-60 Payload Quantity 10 targets Total Activity up to 82,000 Ci

2. Payload Type 2: Type 2 consists of lower-activity targets of an older design, which include:

A. Design in which an aluminum core rod holds pellets placed in dimples on the outer surface and which are retained by a close-fitting outer sleeve, welded to the core rod on each end and B. Design using a solid rod of cobalt inside a stainless steel tube with welded ends.

Table 1.9 includes the characteristics of payload type 2 of the isotope production targets.

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 9341 8 71-9341 USA/9341/B(U)F-96 10 OF 12 5.(b)(1) Type and form of material (continued)

Table 1.9. Characteristics of Isotope Production Targets, Payload Type 2 Parameter Value Target Diameter 5/8 inches (pellet design) 5/16 inches (solid rod design)

Target Length Up to 16.5 inches Cladding Material Aluminum alloy 6061-T6 (pellet design)

Stainless steel (solid rod design)

Target Contents Approximately 5,500 pellets or one solid or segmented rod of cobalt metal Pellet Size 1 mm diameter x 1 mm thick Maximum Activity Up to 4,000 Ci, Co-60 Payload Quantity 20 targets Total Activity Up to 80,000 Ci 5.(b)(2) Maximum quantity of material per package (i) For the contents described in 5(b)(1)(i):

8 irradiated MURR fuel elements. Only one fuel element is allowed per basket location.

(ii) For the contents described in 5(b)(1)(ii):

8 irradiated MITR-II fuel elements. Only one fuel element is allowed per basket location.

(iii) For the contents described in 5(b)(1)(iii):

8 irradiated ATR fuel elements. Only one fuel element is allowed per basket location.

(iv) For the contents described in 5(b)(1)(iv):

19 irradiated TRIGA fuel elements. Only one fuel element is allowed per basket location.

26 types of TRIGA fuel.

(v) For the contents described in 5(b)(1)(v) 8 irradiated PULSTAR fuel elements. Only one fuel element is allowed per basket location.

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 9341 8 71-9341 USA/9341/B(U)F-96 11 OF 12 5.(b)(2) Maximum quantity of material per package (continued)

(vi) For the contents described in 5(b)(1)(vi) 8 irradiated square fuel elements or loose plate boxes. Only one fuel element or loose plate box is allowed per basket location (i.e., compartment). Up to 31 loose plates may be placed in each loose plate box.

(vii) Plutonium Quantity. The maximum quantity of plutonium in the BRR package is 6,500 Ci (at 4% 235U enrichment of PULSTAR fuel).

(viii) For the contents described in 5(b)(1)(vii)(1) 10 target holders. For payload type 1, up to 10 target holders may be placed into the inner row of holes in the isotope basket.

(ix) For the contents described in 5(b)(1)(vii)(2) 20 target holders. For payload Type 2, up to 20 target holders may be placed into any of the 20 holes in the isotope basket.

(c) Criticality Safety Index (CSI): 0

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

(a) Each package shall be operated and prepared for shipment in accordance with Chapter 7 of the application, as supplemented (i) For TRIGA fuel, spacer pedestals shall be used as described in Table 7.1-2 of the application.

(ii) For PULSTAR fuel, spacer pedestals shall be used as described in Table 7.1-1 of the application.

(iii) For square fuel and loose plates, spacer pedestals shall be used as described in Table 7.1-1 of the application.

(iv) When shipping loose plates, use aluminum dunnage sheets to reduce the free space between the flat face of the loose plates and the box opening to a value of 1/4 inches or less. The dimensions of the dunnage sheets shall be as shown in Figure 7.1-1 of the application.

(v) For isotope production targets, a personnel barrier shall be used as described in Section 7.1.4 of the application.

(b) Each package shall be acceptance tested and maintained in accordance with Chapter 8 of the application.

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 9341 8 71-9341 USA/9341/B(U)F-96 12 OF 12

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

8. Transport by air of fissile material is not authorized.

9. Revision No. 6 of the certificate may be used until January 31, 2020.

10. Expiration date: January 31, 2020.

REFERENCES AREVA Federal Services LLC application dated June 13, 2019. (Model No. BRR Safety Analysis Report, Revision 15)

Orano Federal Services LLC supplements dated: July 23 and July 30, 2019 (Safety Analysis Report, Revision 16).

FOR THE U.S. NUCLEAR REGULATORY COMMISSION

/RA/

John McKirgan, Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Date: 8/15/19