ML23139A205
| ML23139A205 | |
| Person / Time | |
|---|---|
| Site: | 07109358 |
| Issue date: | 05/19/2023 |
| From: | Shaw D Orano TN Americas, TN Americas LLC |
| To: | Office of Nuclear Material Safety and Safeguards, Document Control Desk |
| Shared Package | |
| ML23139A204 | List: |
| References | |
| E-62349, CoC No. 9358 | |
| Download: ML23139A205 (1) | |
Text
May 19, 2023 E-62349 Orano TN 7160 Riverwood Drive U. S. Nuclear Regulatory Commission Suite 200 Columbia, MD 21046 Attn: Document Control Desk USA One White Flint North Tel: 410-910-6900 11555 Rockville Pike Fax: 434-260-8480 Rockville, MD 20852
Subject:
Response to Request for Additional Information for the Application for Revision 8 of Certificate of Compliance No. 9358 for the Mo del No. TN-LC, Docket No. 71-9358
References:
(1) TN Letter E-62349. Application for Revision 8 of Certificate of Compliance No. 9358 for the Model No. TN-LC, Docket No. 71-9358 dated February 15, 2023 (Agencywide Documents Access and Management System [ADAMS] Package Accession No. ML23046A179)
(2) Letter to Don Shaw (TN) from Pierre Saverot (NRC), Reques t for Additional Information for the Review of the Model No. TN-LC Package, dated May 15, 2023, Docket Number 71-9358, Enterprise Project Identifier (EPID) No. L-2023-LLA-0019 (ADAMS Package Accession No ML23121A133) with attachment Request for Additional Information for the review of the Model No. TN-LC Package Docket No. 71-9358 (ADAMS Package Accession No ML23121A134)
(3) NRC Certificate of Compliance for the Model No. TN-LC, USA/9358/B(U)F-96, Revision 7 In accordance with 10 CFR 71.39, TN Americas LLC (TN) made a su bmission of an application to revise Certificate of Compliance (CoC) No. 9358 for the TN-LC packaging [1]. The NRC submitted a request for additional information (RAI) needed to continue the review of the application [2].
Enclosure 1 provides the responses to the RAI.
Preliminary changed SAR pages are provided as Revision 11B in E nclosure 2. A consolidated SAR Revision 11 will be submitted upon completion of the NRC review.
The changed pages are indicated by Revision 11, 05/23 in the header of the page.
Each changed page includes a revision bar adjacent to the chang ed content and the changes made relating to Revision 11B are gray shaded to di stinguish them
Enclosures transmitted herein contain SUNSI. When separated fr om enclosures, this transmittal document is decontrolled.
E-62349 Document Control Desk Page 2 of 2
from the Revision 11A changes to the SAR. SAR drawing changes are indicated by clouds around the changed areas. A public version of the Revision 11B SAR changed pages with proprietary information redacted is provided for public availab ility as Enclosure 3.
The NRC Electronic Information Exchange (EIE) system is used fo r submission of this application.
Proposed changes to the NRC Certificate of Compliance [3] are a nnotated and provided as.
Certain portions of this submittal include proprietary informat ion. In accordance with 10 CFR 2.390, TN Americas is providing an affidavit (Enclosure 5) requ esting that this proprietary information be withheld from public disclosure.
Should the NRC staff require additional information to support review of this application, please contact Peter Vescovi at 336-420-8325, or by email at peter.ves covi@orano.group.
Sincerely, Digitally signed by Donis Donis Shaw Shaw Date: 2023.05.19 09:52:02 -04'00' Don Shaw Licensing Manager TN Americas LLC
Enclosures:
- 1. RAIs and Responses
- 2. TN-LC Transportation Package Safety Analysis Report Revision 11B Changed Pages (Proprietary Version)
- 3. TN-LC Transportation Package Safety Analysis Report Revision 11B Changed Pages (Public Version)
- 4. Proposed Certificate of Compliance No. 9358, Revision 8 Mark up
- 5. Affidavit Pursuant to 10 CFR 2.390
cc: Pierre Saverot, Senior Project Manager, U.S. Nuclear Regula tory Commission Peter Vescovi, Licensing Engineer, TN Americas LLC Kamran Tavassoli, Project Manager, TN Americas LLC RAIs and Responses Enclosure 1 to E-62349
Structural Review:
Quantify the maximum axial gap that is available between the basket and the cask cavity described in section 1.4.5.2.1 of the SAR based on the TN -LC cask maximum cavity length and the 1FA basket minimum length, or specify the axial gap limits on the drawing, so it can be verified.
In paragraph 4 of section 1.4.5.2.1 of the SAR, Rev. 11A, it states, neither end cap can come off of the basket compartment once the cask is closed because the axial gap between the basket and the cask cavity is smaller than either the thickness of the top end cap plate, or the combined thickness of the bottom end cap plate and its spacers.
The maximum possible axial gap between the basket and the cask cavity cannot be determi ned by the staff since only the minimum length (182.5 inches) of the TN -LC cask cavity is provided on Drawing 65200-71-01, sheet 4, and the maximum cavity length is not specified. Although the 1FA basket length provided on Drawing 65200-71-90 is 181.5 inches nominal, the minimum length is not specified or cannot be determined without knowing fabrication tolerances. The maximum possible axial gap needs to be determined so it can be compared against the minimum cover plate thickness to validate the evaluation provided in the SAR.
This information is needed to satisfy the requirements of 10 CFR 71.31(b), 71.33, and 71.43(f).
Note 7 of SAR Drawing 65200-71-90 has been updated to clearly specify the maximum allowed axial gap (1 inch) between the basket and the cavity.Additionally, Safety Analysis Report (SAR) Appendix 1.4.1 has been updated to reflect that the Revision 8 version of the Drawing 65200-71-90 has changed to Revision 9.
Impact:
SAR Appendix 1.4.1 and SAR Drawing 65200-71 -90 have been revised as described in the response.
Page 1 of 4 RAIs and Responses Enclosure 1 to E-62349
Justify the use of AWS D1.3 code in lieu of AWS D1.6 code for welding of important to safety damaged fuel end cap assemblies.
Note 3 on DWG No. 65200-71-92, Rev. 0B, requires all welding to be performed per D1.3 or ASME Section IX, as applicable, and all visual inspection of welds per AWS D1.3. The end cap assemblies are made of stainless steel material of thickness varying from 1.25 to 0.105. AWS D1.6 code contains welding requirements for the fabrication of welded stainless steel structures and is intended to cover base metal material thickness of 1/16 inch and above. While AWS 1.3 code contains welding requirements for a typical carbon or alloy sheet steel of material thickness less than or equal to 3/16.
This information is needed to confirm compliance with 10 CFR 71.119.
Note 3 of SAR Drawing 65200-71-92 has been updated to mention American Welding Society (AWS) D1.6 instead of AWS D1.3.
Impact:
SAR Drawing 65200-71-92 has been revised as described in the response.
Page 2 of 4 RAIs and Responses Enclosure 1 to E-62349
Clarify Note 7 on Drawing 65200-71-90, sheet 1, about fuel gap.
In Note 7 on Drawing 65200-71-90 R8, sheet 1, it states, The fuel gap shall not exceed 1.0 inch. This note does not specify in what direction and between which components this gap needs to be verified. This note needs to be clarified to implement this requirement properly.
This information is needed to determine compliance with 10 CFR 71.55(d) and 10 CFR 71.55(e).
Note 7 of SAR Drawing 65200-71-90 has been updated to clearly state the direction of the gap, as well as between which components this gap needs to be verified.
Impact:
SAR Drawing 65200-71-90 has been revised as described in the response.
Page 3 of 4 RAIs and Responses Enclosure 1 to E-62349
a) Justify quality classification of the studs (Item 3) and the nuts (Item 7) shown on Drawing 65200-71-90, sheet 1, as not important to safety (NITS).
The studs and nuts connect the basket tube structure (Items 1 & 2) to the rails (Items 5 &
- 11) that fill the space between the cask wall and the tube structure; while the poison plate (Item 4) and insert plate (Item 8) are sandwiched between the rail and the tube structure. All these components are classified as important to safety (ITS), quality category A, while the studs and nuts that keep these components together in their position inside the cask are classified as NITS. Following guidance in NUREG/CR -6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, it appears to the staff that these components should be classified as ITS Category C Items.
b) Provide the size, material specification, and installation requirements of the studs and nuts (Items 3 & 7) and clarify how they are attached to the basket structure.
The stud size, material specification, and installation requirements are not provided on the drawing. As currently depicted on the drawing, it is not clear how studs are attached to the basket structure (i.e., welded on to the surface of Items 1 & 2 or mechanically fastened with a through thickness holes in Items 1 & 2). It is important to know this for the transport of the damaged fuel, specifically i f the studs are mechanically fastened with a through thickness holes in the basket walls.
This information is needed to satisfy the requirements of 10 CFR 71.33 and 10 CFR 71.51.
Item a:
SAR Drawing 65200-71-90 has been revised to change the quality category of the studs (Item
- 3) and the nuts (Item 7) to ITS Category C.
Item b:
SAR Drawing 65200-71-90 has been revised to provide the size and material specification of the studs and nuts (Items 3 and 7), and to clarify on detail B that the studs are welded to the basket plates.
Impact:
SAR Drawing 65200-71-90 has been revised as described in the response.
Page 4 of 4 Enclosure 2 to E-62127
TN-LC Transportation Package Safety Analysis Report Revision 11B Changed Pages (Proprietary Version)
Withheld Pursuant to 10 CFR 2.390
Enclosure 3 to E-62349
TN-LC Transportation Package Safety Analysis Report Revision 11B Changed Pages (Public Version)
TN-LC Transportation Package Safety Ana lysis Report Revision 11, 05/23
Appendix 1.4.1 TN-LC Transport Package Drawings
Drawing Number Title 65200- 71- 01 Revision 10 TN-LC Cask Assembly (11 sheets) 65200- 71-20 Revision 5 TN-LC Impact Limiter Assembly (2 sheets) 65200- 71-21 Revision 2 TN-LC Transport Packaging Transport Configuration (1 sheet)
Drawing Number Title 65200- 71-40 Revision 4 TN-LC-NRUX Basket Basket Assembly (5 sheets) 65200- 71-50 Revision 4 TN-LC-NRUX Basket Basket Tube Assembly (5 sheets)
Drawing Number Title 65200- 71-60 Revision 4 TN-LC-MTR Basket General Assembly (4 sheets) 65200- 71-70 Revision 4 TN-LC-MTR Basket Fuel Bucket (2 sheets)
Drawing Number Title 65200- 71-80 Revision 4 TN-LC-TRIGA Basket (5 sheets)
Drawing Number Title 65200- 71-90 Revision 9 TN-LC-1FA Basket (5 sheets) 65200- 71-92Revision 0 TN-LC-1FA PWR Basket Damaged Fuel End Caps (1 sheet) 65200- 71-96 Revision 5 TN -LC -1FA BWR Sleeve and Hold -Down Ring (2 sheets) 65200- 71-102 Revision 7 TN-LC -1FA 21 Pin Can Basket (4 sheets)
TN-LC -0100 1.4.1-1 Proprietary and Security Related Information for Drawing 65200-71-90, Rev. 9 Withheld Pursuant to 10 CFR 2.390 Proprietary and Security Related Information for Drawing 65200-71-92, Rev. 0 Withheld Pursuant to 10 CFR 2.390 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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 1 OF 36
- 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.
This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or
- b. 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 TN Americas LLC TN-LC Transportation Package Safety Analysis 7160 Riverwood Drive, Suite 200 Report, Revision No. 10, dated April 2022.
Columbia, MD 21046
- 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.: TN-LC
(2) Description
The packaging, designed for transport of irradiated test, research, and commercial reactor fuel in either a closed transport vehicle or an ISO container, consists of a payload basket, a shielded body, a shielded closure lid and top and bottom impact limiters. The packaging body is a right circular cylinder, approximately 197.5 inches long and 30 inches in diameter, composed of top and bottom end flange forgings connected by inner and outer shells. Lead shielding, made of ASTM B29 copper lead, is placed between the two cylindrical shells, in the bottom end assembly, and in the lid. Neutron shielding, composed of a borated resin compound inserted into twenty aluminum shield boxes, is set between the outer shell and a 0.25 inch-thick Type 304 stainless steel outer sheet. Two removable trunnions are bolted to the packaging body using eight 1-8UNC bolts for each trunnion. Two pocket trunnions in the bottom flange, used for rotating the package, may also be used for horizontal package lifting.
Impact limiters, with an approximate outside diameter of 66 inches and height of 22.75 inches, consisting of balsa and redwood blocks encased in stainless steel shells, are attached to each end of the packaging during shipment, each with eight 1-8UNC bolts.
Four basket designs are provided for transport of Boiling Water Reactor (BWR), Pressurized Water Reactor (PWR), Mixed Oxide Fuel (MOX), Evolutionary Pressurized Reactor (EPR),
National Research Universal Reactor (NRU), National Research Experimental 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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 2 OF 36
5.(a)(2) Description (Continued)
Reactor (NRX), Material Test Reactor (MTR), and Training, Research, and Isotope General Atomics Reactor (TRIGA) fuel assemblies, fuel elements or fuel rods. The packaging may be loaded or unloaded either in a pool or a hot cell environment. The spent fuel payload is shipped dry in a helium atmosphere. The first fabricated packaging, Unit 1, shall only be loaded with the TN-LC 1FA basket.
Nominal weights and dimensions are as follows:
- Overall length with impact limiters: 230 inches
- Overall length without impact limiters: 197.50 inches
- Cavity length (minimum): 182.50 inches
182.10 inches for Unit 1
- Cavity inner diameter: 18 inches
- Lid thickness: 7.50 inches
- Weight of contents: 7,100 lbs
- Weight of lid: 1,000 lbs
- Weight of impact limiters: 3,000 lbs
- Total loaded weight of the package: 51,000 lbs
(3) Drawings
The packaging is constructed and assembled in accordance with the following drawings:
65200-71-01 Revision 10 TN-LC Cask Assembly (11 sheets)
65200-71-20 Revision 5 TN-LC Impact Limiter Assembly (2 sheets) 65200-71-21 Revision 2 TN-LC Transport Packaging Transport Configuration (1 sheet) 65200-71-40 Revision 4 TN-LC-NRUX Basket Basket Assembly (5 sheets) 65200-71-50 Revision 4 TN-LC-NRUX Basket Basket Tube Assembly (5 sheets)
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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 3 OF 36
65200-71-60 Revision 4 TN-LC-MTR Basket General Assembly (4 sheets) 65200-71-70 Revision 4 TN-LC-MTR Basket Fuel Bucket (2 sheets) 65200-71-80 Revision 4 TN-LC-TRIGA Basket (5 sheets)
65200-71-90 Revision 7 TN-LC-1FA Basket (5 sheets) 65200-71-96 Revision 5 TN-LC-1FA BWR Sleeve and Hold-Down Ring (2 sheets) 65200-71-102 Revision 7 TN-LC-1FA Pin Can Basket (5 sheets)
65200-71-91 Revision 0 TN-LC-1FA PWR Basket Damaged Fuel Assembly Can (3 Sheets)
5.(b) Contents
(1) Type and Form of Material
(i) Intact or damaged NRU and NRX Mk I fuel assemblies which meet the specifications listed in Table 1 below, respectively, are authorized for transportation in the TN-LC-NRUX basket.
Intact fuel assemblies are fuel assemblies containing fuel rods with no known or suspected cladding defects greater than hairline cracks or pinhole leaks.
Damaged fuel assemblies, with cladding damage in excess of pin hole leaks or hairline cracks, are authorized only if the total surface area of the damaged cladding does not exceed 5% of the total surface area of each rod.
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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 4 OF 36
5.(b)(1) Type and Form of Materials (continued)
Table 1
NRU and NRX Mk I Fuel Specifications for Transport in the TN-LC-NRUX Basket
Parameter NRU NRX Mk I
Physical and Material Description Number of Assemblies 26 26 Number of rods/assembly 12 7 Assembly length (inch) (1) 116 116 Nominal Assembly mass (g) 4660 5780 Fuel form U-Al U-Al 235U per rod (g) 45.4 75.2 Enrichment (wt.% 235U) 93 93 Cladding and Spacer Material Al Al Thermal and Radiological Parameters Cooling Time (years) (2) 10 10 Depletion (wt.% 235U) (3) 80 80 Decay Heat per Assembly (watts) (4) 15 15
Notes:
- 1. Maximum length of the fuel assembly (unirradiated) for shipment.
- 2. The cooling time of the fuel assembly rounded down to 0.5 years.
- 3. The depletion (or burnup) of the fuel assembly rounded up to 0.5%.
- 4. The decay heat of the fuel assembly is less than 15 watts at the maximum burnup and minimum cooling time.
(ii) Intact or damaged MTR fuel elements that are enveloped or bounded by the fuel element design characteristics listed in Table 2 below, with an average burnup and minimum cooling time as specified in Table 3 below, and a maximum decay heat of 25 watts per element, are authorized for transportation in the TN-LC-MTR basket.
Intact fuel elements are fuel elements containing fuel plates with no known or suspected cladding defects greater than hairline cracks or pinhole leaks.
Damaged fuel elements, with cladding damage in excess of pin hole leaks or hairline cracks, are authorized only if the total surface area of the damaged cladding does not exceed 5% of the total surface area of each element.
The MTR fuel assemblies shall meet all the requirements in Table 3.
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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 5 OF 36
5.(b)(1) Type and Form of Materials (continued)
Table 2
MTR Fuel Element Design Characteristics
Fuel Element Class M-01 M-02 M-03 M-04 M-05 M-06 M-07 M-08(1)
Number of Fuel Plates (2)2321 19 17 10 18 17 23 235U mass per Plate (g) 16 16.5 17.5 19 22 20.5 11.5 22 Active Fuel Width (cm) 6.7 6.7 6.7 6.7 6.7 5.9 6.7 6.7 Active Fuel Length (cm) 56 56 56 56 56 56 27.5 56 Enrichment (wt.% 235U) 94 94 94 94 94 94 94 94 Fuel Element Depth (cm) 7.5 7.5 7.5 7.5 7.5 7.5 7.5 7.5
Notes:
- 1. The M-08 Element class requires that the central stack of fuel elements remain empty. Also, the total 235U mass is limited by the maximum value in Table 3.
- 2. The plate thickness is greater than 0.12 cm and the clad thickness is greater than 0.02 cm.
Table 3 MTR Fuel Element Qualification Burnup Cooling Time Enrichment Type (MWd/MTU) (days) 66,000 740 Type A 165,000 1120 235U Enrichment 90% 330,000 1440 235U Mass 380 g 495,000 1680 660,000 1950 57,750 770 Type B 144,375 1150 235U Enrichment 90% 288,750 1470 380 g < 235U Mass 460 g 433,125 1710 577,500 1950 29,330 740 Type C 73,325 1120 40% 235U Enrichment < 90% 146,650 1440 235U Mass 380 g 219,975 1690 293,300 1940 13,930 830 Type D 34,825 1220 19% 235U Enrichment < 40% 69,650 1560 235U Mass 470 g 104,475 1850 139,300 2150 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 NUMBERc. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 6 OF 36
5.(b)(1) Type and Form of Materials (continued)
Notes Burnup = fuel element average burnup.
Use burnup (MWd/MTU) and Enrichment Type (A, B, C, or D with limits on 235U enrichment and 235U mass per element) to look up minimum cooling time in days.
Licensee is responsible for ensuring that uncertainties in burnup, enrichment, and mass are applied conservatively.
Fuel with burnups greater than those listed for each Enrichment Type is not authorized for transport.
Burnups may be either rounded up to the next higher burnup or linear interpolation may be used to determine the minimum cooling time. However, for conservatism, an additional cooling time of 30 days must be added to any linearly interpolated value.
Example: An M-06 class element with an enrichment of 45 wt.% 235U and a 235U mass of 350 grams is classified as enrichment Type C. A burnup of 100,000 MWd/MTU is acceptable for transport after 1440 days cooling time as defined by 146,650 MWd/MTU from the qualification table (when linear interpolation is not employed). When linear interpolation is employed the minimum required cooling time is 1267 days (1237 days based on interpolation + 30 days additional cooling time).
(iii) Intact TRIGA fuel assemblies/elements that are enveloped by the fuel assemblies/element design characteristics listed in Table 4, intact TRIGA fuel follower control rods that are enveloped by the fuel assembly/element design characteristics listed in Table 5, with an average burnup and minimum cooling time meeting the specifications of Table 6 for fuel assemblies/elements or of Table 7 for follower control rods, and a maximum decay heat of 8 watts per assembly/element, are authorized for shipment with the TN-LC-TRIGA basket.
Intact fuel assemblies/elements are fuel assemblies/elements containing fuel rods with no known or suspected cladding defects greater than hairline cracks or pinhole leaks.
The design characteristics of the TRIGA fuel assemblies/elements are described in Tables 4 and 5 below.
The fuel qualification Tables 6 and 7 specify the maximum assembly/element average burnup and minimum cooling time. The fuel elements/assemblies shall meet all the requirements of Tables 6 and 7.
The poison plates in TN-LC-TRIGA basket are constructed from either boron aluminum alloy, or metal matrix composite (MMC), or Boral. The minimum areal density of Boron-10 (10B) for either the boron enriched aluminum alloy or the metal matrix composite is 5.56 mg/cm2. The minimum areal density of 10B for Boral is 6.67 mg/cm2.
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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 7 OF 36
5.(b)(1) Type and Form of Materials (continued)
Table 4
TRIGA Fuel Assembly/Element Design Characteristics
Assembly/Element Type Al ACPR (1) Standard FLIP (2) FLIP (2) FLIP ( 2)
Clad LEU-I (3) LEU-II (3)
Element ID T-01 T-02 T-03 T-04 T-05 T-06
Fuel Material U-ZrH U-ZrH U-ZrH U-ZrH U-ZrH U-ZrH
Enrichment (wt.% 235U) 20 20 20 70 20 20 235U-Mass (g) 41 56 41 137 101 169 Active Fuel Length (inch) 15 15 15 15 15 15 Pellet Diameter (inch) 1.41 1.41 1.44 1.44 1.44 1.44 Clad Material Al SS304 SS304 SS304 SS304 SS304 H/Zr, max. 1.0 1.7 1.7 1.6 1.6 1.6
Table 5
TRIGA Fuel Follower Control Rods Design Characteristics
Assembly/Element Type Standard FLIP (2) ACPR (1)
LEU-I (3)
Element ID T-07 T-08 T-09
Fuel Material U-ZrH U-ZrH U-ZrH
Enrichment (wt. % 235U) 20 20 20 235U-Mass (g) 38 97 56 Active Fuel Length (inch) 15 15 15 Pellet Diameter (inch) 1.32 1.32 1.32 Clad Material SS304 SS304 SS304 H/Zr, max. 1.7 1.6 1.7
Notes:
- 1. ACPR - Annular Core Pulse Reactor
- 2. FLIP - Fuel Life Improvement Program
- 3. LEU - Low Enriched Uranium 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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 8 OF 36
5.(b)(1) Type and Form of Materials (continued)
Table 6
TRIGA Fuel Qualification for Fuel Assembly/Elements
Burnup Cooling Time Element ID (MWd/MTU) (days)
T-01 35,750 400 71,500 560 107,250 640 143,000 710 T-02 35,750 650 71,500 970 107,250 1310 143,000 1870 T-03 35,750 520 71,500 840 107,250 1170 143,000 1730 T-04 112,500 1000 225,000 1380 337,500 1820 450,000 2520 T-05 35,750 920 71,500 1290 107,250 1710 143,000 2360 T-06 36,500 1190 73,000 1690 109,500 2320 146,000 3170 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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 9 OF 36
5.(b)(1) Type and Form of Materials (continued)
Table 7
TRIGA Fuel Qualification for Fuel Follower Control Rods
Burnup Cooling Time Element ID (MWd/MTU) (days)
T-07 35,750 540 71,500 890 107,250 1280 143,000 1960 T-08 35,750 940 71,500 1350 107,250 1840 143,000 2580 T-09 35,750 670 71,500 1020 107,250 1420 143,000 2100
Notes for Tables 6 and 7:
Burnup = fuel element / assembly / follower control rod average burnup.
Use burnup (MWd/MTU) and Element ID to look-up minimum cooling time in days. Licensee is responsible for ensuring that uncertainties in burnup are applied conservatively.
Fuel with a burnup greater than that listed for each element type in Tables 6 and 7 is unacceptable for transport.
Burnups may be either rounded up to the next higher burnup or linear interpolation may be used to determine the minimum cooling time.
However, for conservatism, an additional cooling time of 30 days must be added to any linearly interpolated value.
Example: A T-03 element with a burnup of 100,000 MWd/MTU is acceptable for transport after 1170 days cooling time as defined by 107,250 MWd/MTU (Table 6, rounding up) on the qualification table (when linear interpolation is not employed). When linear interpolation is employed the minimum required cooling time is 1133 days (1103 days based on interpolation + 30 days additional cooling time).
(iv) Intact or damaged PWR fuel assembly, as specified in Table 8, or intact BWR fuel assembly, as specified in Table 13, or intact or damaged fuel rods in a pin can are authorized for transport with the TN-LC-1FA basket.
Intact fuel assemblies are fuel assemblies containing fuel rods with no known or suspected cladding defects greater than hairline cracks or pinhole leaks.
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 NUMBERc. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 10 OF 36
5.(b)(1) Type and Form of Materials (continued)
Damaged Fuel assemblies have missing or partial-length fuel rods or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks.
The extent of cladding damage is to be limited such that it can be handled by normal means and that a fuel pellet is not able to pass through the gap created by the cladding opening. Damaged fuel assemblies can also contain top and bottom end fittings or nozzles or tie plates depending on the fuel type. Damaged PWR fuel assembly is authorized for transport only when confined in a Fuel Assembly Can (FAC).
Damaged Fuel rods are complete or partial-length fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of cladding damage in the fuel rod is limited such that it can be handled by normal means and that a fuel pellet is not able to pass through the gap created by the cladding opening.
The fuel rods include irradiated PWR, BWR, MOX, and EPR fuel rods. Intact or damaged PWR and intact BWR fuel rods may be from any of the fuel assemblies listed in Table 8 or Table 13, respectively.
MOX rods have the same geometry as PWR or BWR rods, as defined in Table 8 and Table 13. The composition of MOX fuel is specified in Table 12.
The EPR fuel rods are specified in Table 10.
The poison plates in the TN-LC-1FA basket are constructed from boron aluminum alloy, or metal matrix composite (MMC), or Boral. The minimum 10B aeral density of the poison plate is 16.7 mg/cm2 for either the boron aluminum alloy or the MMC. The minimum 10B aeral density of the poison plate is 20.0 mg/cm2 for Boral.
In addition to the poison plates provided in the basket, Poison Rod Assemblies (PRAs) may be used for transportation of PWR fuel assemblies. The minimum required B4C content of the absorber rods in the PRA is 40% Theoretical Density (TD). A summary of the number of absorber rods required in the PRA for each PWR fuel class is shown in Table 11. PRA loading configurations are also illustrated in Figure 1 through Figure 5. Alternatively, in the absence of PRAs, burnup credit restrictions as shown in Table 11a and Table 11b are required for transportation of PWR fuel assemblies. Burnup credit is not applicable to BW 15x15 fuel class.
The PWR fuel assemblies fuel qualification table (FQT) is provided in Tables 15 and 15a. The BWR fuel assemblies FQT is provided in Table 16. The PWR rod FQTs are shown in Table 17 and Table 18 for the 21 and 9 rod configurations, respectively, and in Table 17a for the Unit 1 packaging. The BWR rod FQTs are shown in Table 19 and Table 20 for the 21 and 9 rod configurations, respectively. The MOX rod FQT, provided in Table 21 for both 21 and 9 rods, is applicable to both BWR and PWR MOX rods. The FQTs for the UO2 Standard EPR rods are governed by the PWR rod FQTs (Tables 17, 17a and 18), while the FQT for the MOX EPR rods is governed by the MOX rod FQT (Table 21).
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5.(b)(1) Type and Form of Materials (continued)
Table 8 WR Fuel Specifications for Transport in the TN-LC-1FA Basket
One intact or damaged unconsolidated B&W 17x17, WE Fuel Class (1) (2) 17x17, CE 16x16, B&W 15x15, WE 15x15, CE 15x15, WE 14x14, WE 16x16 or CE 14x14 class PWR assembly (without control components) that are enveloped by the fuel assembly design characteristics listed in Table 9. Reload fuel manufactured by the same or other vendors, but enveloped by the design characteristics listed in Table 9, is also acceptable.
Maximum Assembly + PRA + 1850 lbs. (839 kg) damaged FAC Weight (as applicable)
Fissile Material UO 2
Maximum Initial Uranium Content (4)490 kg/assembly
Maximum Unirradiated Assembly 178.3 inches (4528.8 mm)
Length
Fuel Assembly Average Burnup, Per Tables 15 and 15a Enrichment and Minimum Cooling Time
Maximum Planar Initial Enrichment 5.0 (3) wt.% 235U
Maximum Decay Heat(5) 3.0 kW per Assembly
16.7 mg/cm2 (Natural or Enriched Boron Aluminum Alloy Minimum 10B areal density in poison / Metal Matrix Composite (MMC))
plates 20.0 mg/cm2 (Boral)
Minimum number of absorber rods per Per Table 11 (Use of PRAs is optional except for BW 15x15)
PRA as a function of assembly class Burnup credit Restrictions in the Per Table 11a or 11b absence of PRAs Notes:
- 1. Up to 21 PWR fuel rods from any of the PWR fuel assemblies listed in Table 9 may also be transported in the TN-LC-1FA basket in a 21 pin can. The fuel rods are loaded in a 21 pin can with a cavity length of 169.55 inches (Option 3) which is placed within the TN-LC-1FA basket. The maximum peak burnup for the fuel rods is 90 GWd/MTU. The required cooling time, as a function of a PWR fuel rod burnup and enrichment, is provided in Table 17 or 17a for 21 rods and Table 18 for 9 rods, respectively.
- 2. Up to 21 EPR fuel rods from any of the fuel class listed in Table 9 and meeting EPR rod parameters specified in Table 10 may also be loaded in the TN-LC-1FA basket. The fuel rods are loaded in a 21 pin can with a cavity length of 180.24 inches (Option 1 and Option 2) which is placed within the TN-LC-1FA basket. The maximum peak burnup for the fuel rods is 90 GWd/MTU. The required cooling time, as a function of an EPR fuel rod burnup and enrichment, is provided in Tables 17 or 17a for 21 rods and Table 18 for 9 rods, respectively.
- 3. For CE 15x15, the maximum planar average initial enrichment is 3.60 wt.% 235U.
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5.(b)(1) Type and Form of Materials (continued)
- 4. The maximum initial uranium content is based on the shielding analysis. The listed value is higher than the actual.
- 5. The maximum decay heat per rod is 220 watts when loading up to 9 rods. The maximum decay heat per rod is 120 watts when loading 10 or more (up to 21) rods.
Table 9
PWR Fuel Assembly Design Characteristics for Transportation in the TN-LC-1FA Basket
B&W B&W WE CE WE CE WE CE WE Assembly Class 15x15 17x17 17x17 15x15 15x15 14x14 14x14 16x16 16x16
Maximum Number of Fuel Rods 208 264 264 216 204 176 179 236 235
Maximum Number of Guide/Instrument Tubes 17 25 25 9 21 5 17 5 21 Rod Pitch(1) (inch) 0.568 0.502 0.496 0.550 0.563 0.580 0.556 0.506 0.496 Pellet Diameter(1) (inch) 0.374 0.323 0.323 0.360 0.367 0.382 0.368 0.326 0.323 Clad Outer Diameter(1) 0.416 0.379 0.360 0.417 0.422 0.440 0.400 0.374 0.360 (inch)
Clad Thickness(1) (inch) 0.024 0.024 0.022 0.026 0.024 0.026 0.022 0.023 0.022
Note 1. The fuel assembly fabrication documentation may be used to demonstrate compliance with these parameters which are design nominal values. Maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a fuel assembly class or an array type.
Table 10
Irradiated EPR Fuel Rod Parameters
Parameter Value Maximum Unirradiated Length 179.5 inches Cladding Thickness Nominal 0.022 inch Maximum Initial Uranium Content 2.05 kgU/rod 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) Type and Form of Materials (continued)
Table 11 Summary of PRA Requirements for PWR Fuel Assembly Classes
Diameter of Minimum B4C Number of Absorber Rods in B4C Absorber Content (g/cm)
Assembly Class PRAs and Locations (cm)
WE 17x17 8, Per Figure 4 0.88 0.613 CE 16x16 5, All Guide Tubes 1.02 0.824 BW 15x15 8, Per Figure 3 0.88 0.613 CE 15x15 1, Center Guide Tube 0.76 0.475 WE 15x15 8, Per Figure 2 0.88 0.613 CE 14x14 5, All Guide Tubes 1.02 0.824 WE 14x14 / 0.613 WE 16x16 8, Per Figure 1 / 5 0.88 / 0.68 BW 17x17 8, Per Figure 4 0.76 0.475
Table 11a Maximum Planar Average Initial Enrichment/Minimum Burnup Combination - PWR Fuel Assembly Classes WE 17x17, WE 16x16, WE 15x15, CE 14x14, CE 15x15 and CE 16x16 Fuel Assembly Classes Fresh Fuel 2.90 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)
(GWd/MTU) 5 3.04 3.05 3.06 3.08 10 3.37 3.40 3.42 3.44 15 3.66 3.70 3.74 3.76 20 4.43 4.53 4.61 4.65 25 4.87 5.00 5.00 5.00
WE 14x14 Fuel Assembly Class Fresh Fuel 2.95 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)
(GWd/MTU) 5 3.26 3.26 3.27 3.28 10 3.65 3.65 3.66 3.68 15 3.92 3.96 4.00 4.03 20 4.67 4.80 4.86 4.93 25 5.00 5.00 5.00 5.00 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) Type and Form of Materials (continued)
Table 11b Maximum Planar Average Initial Enrichment/Minimum Burnup Combination - PWR Fuel Assembly Classes - With Control Rod Insertion (1)
WE 17x17, WE 16x16, WE 15x15, CE 14x14, CE 15x15 and CE 16x16 Fuel Assembly Classes Fresh Fuel 2.90 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)
(GWd/MTU) 5 2.97 2.99 3.00 3.01 10 3.29 3.31 3.34 3.36 15 3.54 3.60 3.64 3.66 20 4.21 4.38 4.45 4.53 25 4.75 4.91 4.98 5.00 30 5.00 5.00 5.00
WE 14x14 Fuel Assembly Class Fresh Fuel 2.95 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)
(GWd/MTU) 5 3.20 3.20 3.21 3.23 10 3.57 3.57 3.59 3.59 15 3.81 3.86 3.90 3.90 20 4.48 4.62 4.71 4.78 25 5.00 5.00 5.00 5.00
(1) Fuel assemblies with accumulated control rod insertion through the first 15 GWd/MTU. Fuel assemblies with accumulated control rod insertion greater than the first MTU are not authorized.
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5.(b)(1) Type and Form of Materials (continued)
Table 12 MOX Fuel Rods Specifications for Transport in the TN-LC-1FA Basket PHYSICAL PARAMETERS:
Up to 21 PWR MOX fuel rods with physical parameters as those listed in Table 8.
Up to 21 BWR MOX fuel rods with physical parameters as those listed in Table 13.
Up to 21 EPR MOX fuel rods with physical parameters as those listed in Table 10.
Fissile Material UO2, PuO2 (Mixed Oxide or MOX)
Heavy Metal (HM) Content 2.5 kgU/rod CRITICALITY PARAMETERS:
235U Content in UO2 : 0.5 235U 0.7 wt.%
Plutonium Content: Pu / (U + Pu) 7.0 wt.%
Initial MOX composition Initial 239Pu Content in PuO2 60.0 wt.%
Initial 241Pu Content in PuO2 7.5 wt.%
THERMAL/RADIOLOGICAL PARAMETERS:
238Pu / 239Pu 4.0 wt.%
Initial MOX Composition for Fuel Qualification 239Pu/ PuO2 50 wt.%
241Am / PuO2 70 ppm 235U/U 0.5 wt.%
Burnup and Minimum cooling time for MOX Per Table 21.
rods
Maximum Decay heat per 25 pin can 2.5 kW for the pin can with up to 21 rods 1.8 kW for the pin can with up to 9 rods
16.7 mg/cm2 Boron Aluminum Alloy / Metal Matrix Minimum 10B aeral density in poison plates Composite (MMC) 20.0 mg/cm2 (Boral)
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5.(b)(1) Type and Form of Materials (continued)
Table 13 BWR Fuel Specification for Transport in the TN-LC-1FA Basket
PHYSICAL PARAMETERS: One intact 7x7, 8x8, 9x9, or 10x10 BWR Fuel Class(1) assembly manufactured by General Electric or Exxon/ANF or FANP or ABB or reload fuel manufactured by same or other vendors that are enveloped by the fuel assembly design characteristics listed in Table 14.
Fuel may be transported with or without Channels channels, channel fasteners, or finger springs.
Fissile Material UO 2
Maximum Assembly Weight with Channels 790 lbs
Maximum Unirradiated Assembly Length 176.6 inches
THERMAL/RADIOLOGICAL PARAMETERS:
Maximum Planar Average Initial Enrichment 5.0 wt.% 235U
Fuel Assembly Average Burnup, Enrichment and Minimum Per Table 16.
Cooling Time
Maximum Decay Heat(2) 2.0 kW per Assembly
16.7 mg/cm2 Boron Aluminum Alloy /
Minimum 10B aeral density in poison plates Metal Matrix Composite (MMC) 20.0 mg/cm2 (Boral)
Notes:
- 1. Up to 21 fuel rods from any of the BWR fuel assemblies listed in Table 14 may also be transported in the TN-LC-1FA basket in the 21 pin can. The fuel rods are loaded in a 21 pin can with a cavity length of 169.55 inches which is placed within the TN-LC-1FA basket. The required cooling time as a function of BWR fuel rod burnup and enrichment are provided in Table 19 for 21 rods and Table 20 for 9 rods, respectively.
- 2. The maximum decay heat per rod is 220 watts when loading up to 9 rods. The maximum decay heat per rod is 120 watts when loading 10 or more (up to 25) rods.
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5.(b)(1) Type and Form of Materials (continued)
Table 14 BWR Fuel Assembly Design Characteristics(1) for Transportation in the TN-LC-1FA Basket (Part 1 of 3)
Transnuclear ID 7x7-49/0 8x8-63/1 8x8-62/2 8x8-60/4 8x8-60/1 9x9-74/2 GE1 GE4 GE-5 GE8 Type II GE9 GE11 Initial Design or GE2 GE-Pres GE10 GE13 Reload GE3 GE-Barrier Fuel Designation GE8 Type I
FANP 8x8-2 Maximum Number of 49 63 62 60 60 74 Fuel Rods Maximum Initial Uranium Content 198 192 192 192 192 192 (kg)
Rod Pitch(5) (inch) 0.738 0.640 0.640 0.640 0.640 0.566 Pellet Diameter(5) 0.487 0.416 0.411 0.411 0.411 0.376 (inch)
Clad Outer 0.563 0.493 0.483 0.483 0.483 0.440 Diameter(5) (inch)
Clad Thickness(5) 0.032 0.034 0.032 0.032 0.032 0.028 (inch)
Table 14 BWR Fuel Assembly Design Characteristics(1) for Transportation in the TN-LC-1FA Basket (Part 2 of 3)
Transnuclear ID 10x10- 7x7-7x7-8x8-FANP 9x9 Siemens 92/2 49/0Z 48/1Z 60/4Z QFA GE12 ENC-IIIA ENC-III(2) ENC Va FANP9 9x9 Initial Design or Reload 9x9 (3)
Fuel Designation GE14 ENC Vb
Maximum Number of 92 49 48 60 81 72 Fuel Rods
Maximum Initial 192 198 198 192 192 192 Uranium Content (kg)
Rod Pitch (5) (inch) 0.510 0.738 0.738 0.642 0.572 0.570 Pellet Diameter(5) (inch) 0.345 0.488 0.491 0.420 0.357 0.374 Clad Outer Diameter(5) 0.404 0.570 0.570 0.501 0.424 0.433 (inch)
Clad Thickness(5) (inch) 0.026 0.035 0.035 0.036 0.030 0.026 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) Type and Form of Materials (continued)
Table 14 BWR Fuel Assembly Design Characteristics(1) for Transportation in the TN-LC-1FA Basket (Part 3 of 3)
Transnuclear ID 10x10-91/1 ABB-8x8 ABB-10x10 LaCrosse ATRIUM 10 SVEA-64 SVEA-100(4) Allis Chalmers -
Initial Design or Reload 10x10
Fuel Designation ATRIUM 10XM Exxon/ANF 10x10
Maximum Number of 91 64 100 100 Fuel Rods
Maximum Initial Uranium 192 192 192 125 Content (kg)
Rod Pitch (5) (inch) 0.510 0.622 0.512 0.565 Pellet Diameter(5) (inch) 0.350 0.411 0.346 0.350 Clad Outer Diameter(5) 0.405 0.378 0.378 0.394 (inch)
Clad Thickness(5) (inch) 0.023 0.024 0.022 0.020
Notes:
- 1. Any fuel channel average thickness up to 0.120 inch is acceptable on any of the fuel designs.
- 2. Includes ENC-IIIE and ENC-IIIF.
- 4. Includes SVEA-92, SVEA-96, SVEA-96+, SVEA-96 OPTIMA, SVEA-96 OPTIMA2, SVEA-96+/L.
- 5. The fuel assembly fabrication documentation may be used to demonstrate compliance with these fuel assembly parameters. The fuel assembly parameters are design nominal values. The maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a fuel assembly class (or an array type).
(2) Maximum quantity of material per package
(i) For the contents described in Item 5(b)(1)(i): 26 intact or damaged either NRU or NRX Mk I fuel assemblies, with an approximate maximum payload of 331 lb.
(ii) For the contents described in Item 5(b)(1)(ii): 54 intact or damaged MTR fuel elements, with an approximate maximum payload of 1,620 lb.
(iii) For the contents described in Item 5(b)(1)(iii): 180 intact TRIGA fuel elements/assemblies with an approximate maximum payload of 2,380 lb.
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5.(b)(2) Maximum quantity of material per package (continued)
(iv) For the contents described in Item 5(b)(1)(iv): one intact PWR fuel assembly, one damaged PWR fuel assembly confined in a FAC, or one intact BWR fuel assembly, or up to 21 intact or damaged PWR (including intact MOX and EPR) or intact BWR fuel rods in a pin can. When transporting 9 or fewer fuel rods, the rods shall be placed in the center 3x3 region of the pin can. The approximate maximum payload is 1,850 lb per PWR assembly with PRAs and Fuel Assembly Can (as applicable), 790 lb per BWR assembly with channels, and 16 lb per fuel rod.
(v) For the Unit 1 packaging, contents described in Item 5(b)(1)(iv) are limited to: one intact PWR fuel assembly, one damaged PWR fuel assembly confined in a FAC, or up to 21 intact or damaged PWR (excluding intact MOX and EPR) fuel rods in a pin can. When transporting 9 or fewer fuel rods, the rods shall be placed in the center 3x3 region of the pin can. The approximate maximum payload is 1,850 lb per PWR assembly with PRAs and FAC as applicable, and 16 lb per fuel rod.
(3) The maximum decay heat for any payload is 3.0 kW.
5(c) Criticality Safety Index (CSI):
For NRU and NRX fuel assemblies described in 100 5(b)(1)(i) and limited in 5(b)(2)(i)
For MTR fuel elements described in 100 5(b)(1)(ii) and limited in 5(b)(2)(ii)
For TRIGA fuel assemblies/elements described in 0 5(b)(1)(iii) and limited in 5(b)(2)(iii)
For intact BWR fuel assemblies described in 0 5(b)(1)(iv) and limited in 5(b)(2)(iv)
For PWR fuel assemblies described in 100 5(b)(1)(iv) and limited in 5(b)(2)(iv) and 5(b)(2)(v)
For fuel rods in a 21 pin can described in 0 5(b)(1)(iv) and limited in 5(b)(2)(iv) and (5(b)(2)(v)
USA/9358/B(U)F-96 Page 20 of 36
Table 15 Fuel Qualification Table for a PWR Fuel Assembly (Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 2.25 2.25 2.20 2.10 2.05 2.05 2.05 2.00 2.00 2.00 2.00 2.00 2.00 1.95 1.95 1.95 1.95 1.95 1.95 1.95 1.95 1.95 1.95 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 1.90 20 3.37 3.35 3.30 3.20 3.05 2.90 2.90 2.85 2.85 2.80 2.80 2.80 2.75 2.75 2.75 2.75 2.75 2.75 2.70 2.70 2.70 2.70 2.65 2.65 2.65 2.65 2.65 2.65 2.60 2.60 2.60 2.60 2.60 2.60 2.60 2.55 30 4.70 4.35 4.10 3.80 3.70 3.65 3.60 3.60 3.55 3.50 3.45 3.45 3.40 3.35 3.35 3.35 3.30 3.30 3.25 3.25 3.20 3.20 3.15 3.15 3.15 3.15 3.15 3.15 3.10 3.10 3.10 3.05 3.05 3.05 39 4.95 4.85 4.75 4.65 4.55 4.45 4.40 4.35 4.25 4.20 4.15 4.10 4.00 3.95 3.95 3.90 3.85 3.80 3.75 3.70 3.70 3.70 3.65 3.65 3.60 3.55 3.55 3.50 3.50 3.50 3.50 40 4.55 4.45 4.35 4.30 4.25 4.15 4.15 4.10 4.05 4.00 3.90 3.90 3.90 3.85 3.80 3.75 3.70 3.70 3.65 3.65 3.65 3.60 3.55 3.55 3.50 45 5.40 5.25 5.15 5.05 4.95 4.85 4.80 4.70 4.60 4.55 4.50 4.45 4.35 4.35 4.30 4.20 4.15 4.10 4.10 4.05 4.00 3.95 3.95 3.90 3.85 50 6.80 6.60 6.50 6.25 6.15 6.00 5.85 5.75 5.60 5.50 5.40 5.30 5.20 5.10 5.05 4.95 4.90 4.85 4.75 4.70 4.65 4.55 4.55 4.50 4.40 55 8.85 8.60 8.30 8.05 7.85 7.65 7.35 7.15 7.00 6.80 6.65 6.45 6.30 6.20 6.05 5.90 5.85 5.70 5.65 5.50 5.45 5.35 5.30 5.25 5.15 60 11.55 11.2010.8510.50 10.159.80 9.55 9.20 8.95 8.70 8.45 8.25 8.00 7.80 7.55 7.40 7.20 7.05 6.85 6.75 6.60 6.45 6.35 6.25 6.10 61 12.15 11.8011.4511.10 10.7010.3510.10 9.75 9.45 9.20 8.90 8.65 8.35 8.20 7.90 7.75 7.55 7.40 7.20 7.00 6.85 6.75 6.55 6.50 6.40 62 12.80 12.4012.0511.65 11.3010.9010.6510.25 9.95 9.70 9.40 9.10 8.85 8.55 8.35 8.15 7.90 7.70 7.50 7.30 7.20 7.05 6.85 6.75 6.65
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 21 of 36
Table 15a Fuel Qualification Table for a PWR Fuel Assembly - Unit 1 Packaging (Minimum required years of cooling time after reactor core discharge)
Burn-u p, Enrichment, wt. % U-235 GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 3.0 3.0 2.9 2.9 2.8 2.8 2.8 2.8 2.7 2.7 2.7 2.7 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.4 2.4 2.4 2.4 2.4 20 4.7 4.6 4.5 4.4 4.3 4.2 4.2 4.1 4.0 4.0 3.9 3.9 3.8 3.8 3.8 3.7 3.7 3.6 3.6 3.6 3.6 3.5 3.5 3.5 3.5 3.4 3.4 3.4 3.4 3.4 3.4 3.3 3.3 3.3 3.3 3.3 30 6.7 6.5 6.3 6.2 6.0 5.9 5.7 5.6 5.5 5.3 5.2 5.1 5.1 5.0 4.9 4.8 4.7 4.7 4.6 4.6 4.5 4.5 4.4 4.4 4.3 4.3 4.2 4.2 4.2 4.1 4.1 4.1 4.1 4.0 39 7.1 6.9 6.7 6.6 6.4 6.3 6.2 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.8 4.7 4.7 4.6 4.6 4.6 40 6.4 6.2 6.1 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.7 4.7 4.7 50 9.6 9.4 9.2 8.9 8.7 8.5 8.3 8.1 7.9 7.8 7.6 7.5 7.3 7.2 7.1 6.9 6.8 6.7 6.6 6.5 6.4 6.3 6.2 6.2 6.1 55 12. 11. 11. 11. 10. 10. 10. 10. 9.8 9.6 9.4 9.1 8.9 8.8 8.6 8.4 8.2 8.1 7.9 7.8 7.7 7.5 7.4 7.3 7.2 0 7 4 1 8 5 3 0 60 14. 14. 14. 13. 13. 13. 12. 12. 12. 11. 11. 11. 11. 10. 10. 10. 10. 9.8 9.6 9.4 9.3 9.1 8.9 8.8 8.6 8 4 1 7 4 0 7 4 1 8 5 3 0 7 5 3 1 61 15. 15. 14. 14. 13. 13. 13. 12. 12. 12. 12. 11. 11. 11. 10. 10. 10. 10. 10. 9.8 9.6 9.4 9.3 9.1 8.9 4 0 7 3 9 6 3 9 6 3 0 7 5 2 9 7 5 2 0 62 16. 15. 15. 14. 14. 14. 13. 13. 13. 12. 12. 12. 11. 11. 11. 11. 10. 10. 10. 10. 10. 9.8 9.6 9.4 9.3 1 7 3 9 5 2 8 5 1 8 5 2 9 7 4 1 9 7 4 2 0 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Note:
- 1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 22 of 36
Table 16 Fuel Qualification Table for a BWR Fuel Assembly (Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/ 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 MTU 10 0.65 0.65 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 20 0.95 0.95 0.90 0.85 0.80 0.80 0.80 0.80 0.80 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 30 1.25 1.20 1.15 1.10 1.10 1.10 1.10 1.10 1.10 1.10 1.10 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.95 0.95 0.95 0.95 39 1.40 1.40 1.40 1.35 1.35 1.35 1.35 1.30 1.30 1.30 1.30 1.30 1.25 1.25 1.25 1.25 1.25 1.20 1.20 1.20 1.20 1.20 1.20 1.15 1.15 1.15 1.15 1.15 1.15 1.15 1.15 40 1.40 1.40 1.35 1.35 1.35 1.35 1.30 1.30 1.30 1.30 1.30 1.25 1.25 1.25 1.25 1.25 1.25 1.25 1.20 1.20 1.20 1.20 1.20 1.20 1.20 45 1.60 1.60 1.60 1.55 1.55 1.55 1.55 1.50 1.50 1.50 1.50 1.50 1.50 1.50 1.50 1.50 1.45 1.45 1.45 1.45 1.45 1.45 1.40 1.40 1.40 50 1.85 1.85 1.85 1.80 1.80 1.80 1.75 1.75 1.75 1.75 1.75 1.75 1.75 1.75 1.70 1.70 1.70 1.70 1.65 1.65 1.65 1.65 1.65 1.60 1.60 55 2.10 2.10 2.10 2.05 2.05 2.05 2.00 2.00 2.00 1.95 1.95 1.95 1.95 1.95 1.95 1.95 1.90 1.90 1.90 1.90 1.90 1.85 1.85 1.85 1.85 60 2.35 2.35 2.35 2.30 2.30 2.30 2.25 2.25 2.25 2.20 2.20 2.20 2.20 2.20 2.20 2.15 2.15 2.15 2.15 2.10 2.10 2.10 2.10 2.05 2.05 61 2.40 2.40 2.40 2.35 2.35 2.35 2.30 2.30 2.30 2.25 2.25 2.25 2.20 2.20 2.20 2.20 2.20 2.20 2.20 2.15 2.15 2.15 2.15 2.10 2.10 62 2.45 2.45 2.45 2.40 2.40 2.40 2.35 2.35 2.35 2.30 2.30 2.30 2.25 2.25 2.25 2.25 2.25 2.25 2.25 2.20 2.20 2.20 2.20 2.15 2.15
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 23 of 36
Table 17 Fuel Qualification Table for 21 PWR/EPR Fuel Rods (UO2)
(Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/ 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 MTU 10 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 20 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 45 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.30 0.30 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 55 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 60 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 61 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.30 0.30 62 0.40 0.40 0.40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 65 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 70 0.50 0.50 0.50 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 75 0.65 0.65 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.50 0.50 0.50 0.50 80 0.85 0.85 0.75 0.75 0.75 0.75 0.75 0.70 0.70 0.70 0.70 0.70 0.70 0.70 85 1.05 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 90 1.25 1.25 1.25 1.15 1.15 1.15 1.10 1.10 1.10 1.00 1.00 1.00 1.00 0.95
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 24 of 36
Table 17a Fuel Qualification Table for 21 PWR Fuel Rods (UO2) - Unit 1 Packaging (Minimum required years of cooling time after reactor core discharge)
Burn-u p, Enrichment, wt. % U-235 GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.35 0.35 0.300.30 0.300.300.30 0.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 20 0.35 0.35 0.300.30 0.300.300.30 0.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 30 0.300.30 0.300.300.30 0.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 39 0.300.30 0.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 40 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 45 0.300.30 0.300.30 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 50 0.350.35 0.350.35 0.30 0.300.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.30 0.30 0.300.30 0.300.300.30 0.30 55 0.370.37 0.370.36 0.36 0.350.350.35 0.35 0.350.35 0.350.35 0.35 0.350.35 0.350.35 0.35 0.350.35 0.350.350.35 0.35 60 0.500.49 0.480.47 0.47 0.460.450.45 0.44 0.440.43 0.430.42 0.42 0.410.41 0.410.40 0.40 0.400.39 0.390.390.38 0.38 61 0.530.52 0.51 0.5 0.5 0.490.480.47 0.47 0.460.46 0.450.44 0.44 0.440.43 0.430.42 0.42 0.420.41 0.410.410.40 0.40 62 0.560.55 0.540.53 0.53 0.520.51 0.5 0.49 0.490.48 0.480.47 0.46 0.460.45 0.450.44 0.44 0.440.43 0.430.430.42 0.42 65 0.560.55 0.55 0.540.53 0.530.52 0.51 0.510.50 0.500.490.49 0.49 70 0.710.70 0.69 0.680.67 0.670.66 0.65 0.650.64 0.630.620.62 0.61 75 0.870.86 0.85 0.840.83 0.82 0.8 0.79 0.790.78 0.770.760.75 0.75 80 1.041.03 1.01 1.000.99 0.970.96 0.95 0.940.93 0.920.910.90 0.89 85 1.241.22 1.20 1.181.16 1.151.131.12 1.101.09 1.081.061.05 1.04 90 1.471.44 1.41 1.391.37 1.341.32 1.3 1.291.27 1.251.231.22 1.20 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Note1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 25 of 36
Table 18 Fuel Qualification Table for 9 PWR/EPR Fuel Rods (UO2)
(Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 20 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 45 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 55 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 60 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 61 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 62 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 65 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 70 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 75 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 80 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 85 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 90 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
- 1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 26 of 36
Table 19 Fuel Qualification Table for 21 BWR Fuel Rods (UO2)
(Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.250.25 20 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.250.25 30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.300.30 39 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.350.35 40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.350.35 45 0.45 0.45 0.45 0.45 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.400.40 50 0.60 0.60 0.60 0.60 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.50 0.50 0.50 0.50 0.50 0.500.50 55 0.75 0.75 0.75 0.75 0.75 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.650.65 60 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.75 0.75 0.75 0.75 0.75 0.75 0.750.75 61 1.05 1.05 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.75 0.75 0.750.75 62 1.10 1.05 1.05 1.05 1.05 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.850.85 65 1.05 1.05 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.900.90 70 1.20 1.20 1.20 1.15 1.15 1.15 1.15 1.15 1.15 1.10 1.10 1.10 1.101.10 75 1.45 1.45 1.45 1.40 1.40 1.40 1.30 1.30 1.30 1.30 1.25 1.25 1.251.25 80 1.70 1.70 1.65 1.65 1.60 1.60 1.60 1.50 1.50 1.50 1.45 1.45 1.451.45 85 2.15 2.05 2.00 2.00 1.95 1.85 1.85 1.80 1.80 1.70 1.70 1.65 1.651.65 90 2.60 2.55 2.50 2.40 2.35 2.30 2.20 2.15 2.15 2.10 2.00 2.00 1.951.95
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
- 1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 27 of 36
Table 20 Fuel Qualification Table for 9 BWR Fuel Rods (UO2)
(Minimum required years of cooling time after reactor core discharge)
Burnup, Enrichment (wt. % 235U)
GWd/
MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 20 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 45 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 55 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 60 0.30 0.30 0.30 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 61 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.25 0.25 0.25 62 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 65 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 70 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 75 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 80 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 85 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 0.50 90 0.60 0.60 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55
0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0
Notes:
1. Explanatory notes and limitations regarding the use of this table follow Table 21.
USA/9358/B(U)F-96 Page 28 of 36
Table 21
Fuel Qualification Table for MOX PWR/BWR/EPR 21 Rods and MOX PWR/BWR/EPR 9 Rods
9 Rods 25 Rods Burnup, GWd/MTHM 0.5 wt.% of 235U0.7 wt.% of 235U0.5 wt.% of 235U0.7 wt.% of 235U
10 0.25 0.25 0.25 0.25 20 0.25 0.25 0.30 0.30 30 0.25 0.25 0.50 0.50 40 0.25 0.25 0.95 0.95 45 0.25 0.25 1.25 1.25 50 0.35 0.35 1.70 1.70 55 0.40 0.40 2.20 2.10 60 0.45 0.45 2.80 2.70 62 0.55 0.55 3.75 3.65 Notes:
- 1. Explanatory notes and limitation regarding the use of this table are provided on the following page.
USA/9358/B(U)F-96 Page 29 of 36
Notes:
General
- 1. Use burnup and enrichment to look up minimum cooling time in years. Licensee is responsible for ensuring that uncertainties in fuel enrichment and burnup are correctly accounted for during fuel qualification.
- 2. For values not explicitly listed in the tables, round burnups up to the first value shown, round enrichments down, and select the cooling time listed.
- 3. UO2 Fuel with an initial enrichment less than 0.7 (or less than the minimum provided above for each burnup) or greater than 5.0 wt.% 235U is unacceptable for transportation.
- 4. Shaded areas in these Tables indicate fuel is not analyzed for loading.
For Fuel Assemblies
- 1. Burnup = Assembly Average burnup.
- 2. Enrichment = Assembly Average Enrichment.
- 3. Fuel assembly with a burnup greater than 62 GWd/MTU is unacceptable for transportation.
For Fuel Rods
- 4. Burnup = Maximum burnup.
- 5. Enrichment = Rod Average Enrichment.
- 6. When transporting 21 or less fuel rods, the rods shall be placed in a specially designed pin can.
- 7. When transporting 9 or less fuel rods, the rods shall be placed in the 3x3 region of the pin can.
- 8. Fuel rods with a burnup greater than 90 GWd/MTU are unacceptable for transportation.
Example: Per Table 15, a PWR assembly with an initial enrichment of 4.85 wt.% 235U and a burnup of 41.5 GWd/MTU is acceptable for transport after a 3.95-year cooling time as defined by 4.8 wt.% 235U (rounding down) and 45 GWd/MTU (rounding up) on the qualification table (other considerations not withstanding).
USA/9358/B(U)F-96 Page 30 of 36
Poison Rod Locations Empty Guide Tube Locations
Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.
Figure 1 PRA Insertion Locations for WE 14x14 Class Assemblies USA/9358/B(U)F-96 Page 31 of 36
Poison Rod Locations Empty Guide Tube Locations
Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.
Figure 2 PRA Insertion Locations for WE 15x15 Class Assemblies USA/9358/B(U)F-96 Page 32 of 36
Poison Rod Locations Empty Guide Tube Locations
Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.
Figure 3 PRA Insertion Locations for BW 15x15 Class Assemblies USA/9358/B(U)F-96 Page 33 of 36
Poison Rod Locations Empty Guide Tube Locations
Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.
Figure 4 PRA Insertion Locations for BW 17x17 and WE 17x17 Class Assemblies USA/9358/B(U)F-96 Page 34 of 36
Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.
Figure 5 PRA Insertion Locations for WE 16x16 Class Assemblies 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 NUMBERPAGE PAGES 9358 7 71-9358 USA/9358/B(U)F-96 35 OF 36
- 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 of Chapter No. 7 of the application, and
(b) Each packaging must meet the Acceptance Tests and Maintenance Program of Chapter No.
8 of the application.
- 7. Transport by air of fissile material is not authorized.
- 8. Prior to the first shipment, the package shall be tested for the entire containment boundary, e.g., all base metal, all joining containment welds, vent port plug seal, drain port plug seal, lid seal, and bottom plug seal, in accordance with ANSI N14.5-2014, by helium leakage rate testing to meet the leaktight criteria of 1.0x10-7 ref-cm3/sec for fabrication leakage tests.
- 9. Poison Rod Assemblies, required for shipment of PWR assemblies if burnup credit is not considered, shall be installed such that the active fuel length is covered by the absorber, and measures shall be taken against their inadvertent removal from the fuel assembly.
- 10. The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 71.17.
- 11. Revision No. 6 of this certificate may be used until June 30, 2023.
- 12. Expiration date: May 31, 2027.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES
- 1. a. CERTIFICATE NUMBER 9358b. REVISION NUMBER 7c. DOCKET NUMBER 71-9358d. PACKAGE IDENTIFICATION NUMBER USA/9358/B(U)F-96PAGE PAGES 36 OF 36 REFERENCES TN-LC Transportation Package Safety Analysis Report, Revision No. 10, dated April 2022.
Renewal Request letter dated March 22, 2022.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION Date See digital signature Name:
W. Allen, Acting Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Signed by Allen, William on 06/23/22 Enclosure 5 to E-62349 AFFIDAVIT PURSUANT TO 10 CFR 2.390
TN Americas LLC )
State of Maryland ) SS County of Howard )
I, Prakash Narayanan, depose and say that I am Chief Technical Officer of TN Americas LLC, duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below, I am submitting this affidavit in conformance with the provisions of 10 CFR 2.390 of the Commissions regulations for withholding this information.
The information for which proprietary treatment is sought is listed below:
- Enclosure 2 - Portions of certain chapters and appendices of the Safety Analysis Report (SAR) for Certificate of Compliance No. 9358 TN-LC, Revision 1 IB, Docket 71-9358 (Proprietary Version)
This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged, or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commissions regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
- 1) The information sought to be withheld from public disclosure involves certain design details associated with the SAR analyses and SAR drawings for the TN-LC System, which are owned and have been held in confidence by TN Americas LLC.
- 2) The information is of a type customarily held in confidence by TN Americas LLC and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.
- 3) Public disclosure of the information is likely to cause substantial harm to the competitive position of TN Americas LLC because the information consists of descriptions of the design and analysis of a radioactive material transportation system, the application of which provide a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their products position or impair the position of TN America LLCs product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.
Further the deponent sayeth not.
Prakash Narayanan Chief Technical Officer, TN Americas LLC
Subscribed and sworn before me this ' °day of May, 2023.
MARYANNE D ATIENZA Notary Public My Commission Expires Anne Arundel County Maryland My Commission Expires March 02, 2025
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