ML23347A137
| ML23347A137 | |
| Person / Time | |
|---|---|
| Site: | 07109313 |
| Issue date: | 12/18/2023 |
| From: | Yoira Diaz-Sanabria Storage and Transportation Licensing Branch |
| To: | |
| Shared Package | |
| ML23347A136 | List: |
| References | |
| EPID L-2021-LLA-0086, EPID L-2022-RNW-0002 | |
| Download: ML23347A137 (1) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 1
OF 9
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 TN Americas LLC 7160 Riverwood Drive, Suite 200 Columbia, MD 21046 TN-40 Transportation Packaging Safety Analysis Report, Revision 17, dated November 2023
- 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-40 (2)
==
Description:==
For descriptive purposes, all dimensions are approximate nominal values.
Actual dimensions with tolerances are as indicated on the drawings.
The TN-40 is designed to transport up to 40 Pressurized Water Reactor (PWR) spent nuclear fuel assemblies discharged from the Prairie Island Nuclear Generating Plant (PINGP). These assemblies have been stored prior to shipment in the TN-40 package used as a dry storage cask at PINGP under SNM-2506.
These 29 loaded packages at the PINGP are authorized for single use. The TN-40 packaging consists of a basket assembly, a containment vessel, a package body which also functions as the gamma shield and neutron shield, and impact limiters. A transport frame, which is not part of the packaging, is used for tie-down purposes.
The containment vessel components consist of the inner shell and bottom inner plate, shell flange, lid outer plate, lid bolts, penetration cover plates and bolts (vent and drain), and the inner metallic seals of the lid seal and the vent and drain seals. The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. The overall containment vessel length is approximately 170.5 in. with a wall thickness of 1.5 in. The cylindrical cask cavity has a nominal diameter of 72.0 in. and a length of 163 in.
Double metallic seals are used for the lid closure. To preclude air in-leakage, the cask cavity is pressurized with helium above atmospheric pressure. The cask cavity is accessed via draining and venting ports.
Double metallic seals are utilized to seal these two lid penetrations. The over-pressure (OP) port provides access to the volumes between the double seals in the lid and cover plates for leak testing purposes. The OP port cover is not part of the containment boundary.
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 2
OF 9
5.(a)(2)
Description (Continued)
The carbon steel packaging body, which also functions as the gamma shielding, is around the inner shell and the bottom inner plate of the containment vessel. The 8.0 in. and 8.75 in. gamma shield completely surround the containment vessel shell and bottom plate, respectively. A 6.0 in. thick shield plate is also welded to the inside of the 4.5 in. thick lid outer plate.
Radial neutron shielding is provided by a borated polyester resin compound surrounding the gamma shield shell. The total radial thickness of the resin and aluminum is 4.50 in. The array of resin-filled containers is enclosed within a 0.50 in. thick outer steel shell. The aluminum container walls also provide a path for heat transfer from the gamma shield shell to the outer shell. A pressure relief valve is mounted on top of the resin enclosure to limit the possible internal pressure increase under hypothetical accident conditions.
The basket structure consists of an assembly of stainless steel cells joined by a fusion welding process and separated by aluminum and poison plates which form a sandwich panel. The panel consists of two aluminum plates separated by a poison plate. The aluminum plates provide the heat conduction paths from the fuel assemblies to the cask inner plate. The poison material provides the necessary criticality control.
The opening of the cells is 8.05 in. x 8.05 in. which provides a minimum of 1/8 in. clearance around the fuel assemblies. The overall basket length (160.0 in.) is less than the cask cavity length to allow for thermal expansion and fuel assembly handling.
The impact limiters consist of balsa wood and redwood blocks encased in stainless steel plates. The impact limiters have an outside diameter of 144 in., and an inside diameter of 92 in. to accommodate the cask ends.
The bottom limiter is notched to fit over the lower trunnions. The impact limiters are attached to each other using tie rods. The impact limiters are also attached to the outer shell of the cask with bolts. Each impact limiter is provided with fusible plugs that are designed to melt during a fire accident, thereby relieving excessive internal pressure. Each impact limiter has lifting lugs for handling, and support angles for holding the impact limiter in a vertical position during storage. An aluminum spacer is placed on the cask lid prior to mounting the top impact limiter to provide a smooth contact surface between the lid and the top impact limiter.
The nominal external dimensions, with impact limiters, are 261 in. long by 144 in. wide. The total weight of the package is 271,500 pounds (lbs.).
5.(a)(3)
Drawings The packaging is fabricated and assembled in accordance with TN Americas LLC Drawing Nos.:
Drawing No Title 10421-71-1 Rev. 6 TN-40 Transport Packaging Parts List and Notes (1 sheet) 10421-71-2 Rev. 3 TN-40 Transport Packaging Transport Configuration (2 sheets) 10421-71-3 Rev. 3 TN-40 Transport Packaging General Arrangement (1 sheet) 10421-71-4 Rev. 0 TN-40 Transport Packaging Lid Assembly and Details (1 sheet)
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 3
OF 9
5.(a)(3)
Drawings (Continued) 10421-71-5 Rev. 0 TN-40 Transport Packaging Lid Details (1 sheet) 10421-71-6 Rev. 0 TN-40 Transport Packaging Trunnion, Basket Rail and Neutron Shield Details (1 sheet) 10421-71-7 Rev. 3 TN-40 and TN-40HT Transport Packaging Impact Limiter Spacer Details (1 sheet) 10421-71-8 Rev. 0 TN-40 Transport Packaging Basket Assembly (1 sheet) 10421-71-9 Rev. 0 TN-40 Transport Packaging Basket Details (1 sheet) 10421-71-40 Rev. 2 TN-40 Transport Packaging Impact Limiters General Arrangement (1 sheet) 10421-71-41 Rev. 2 TN-40 and TN-40HT Transport Packaging Impact Limiters Parts List and Notes (1 sheet) 10421-71-42 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Assembly (1 sheet) 10421-71-43 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Details (1 sheet) 10421-71-44 Rev. 1 TN-40 and TN-40HT Transport Packaging Impact Limiters Parts (1 sheet) 5.(b)
Contents (1)
Type, form, and quantity of material The characteristics of the contents of the TN-40 packaging are limited to the following.
I.
Fuel shall be unconsolidated.
II.
Fuel shall be limited to the following fuel types with specifications depicted in Table 1-1 of this certificate:
i.
Exxon 14X14 Standard, ii.
Exxon 14x14 High Burnup, iii. Exxon 14X14 TOPROD, iii. Westinghouse (WE) 14X14 Standard, and iv. Westinghouse 14X14 OFA.
III. Fuel shall only have been irradiated at the PINGP Unit 1, cycles 1 through 16 or Unit 2, cycles 1 through 15.
IV. The fuel assemblies from Unit 1, Region 4, i.e., assemblies identified as D-01 through D-40, are not authorized contents.
V. Fuel may include burnable poison rod assemblies (BPRAs) provided:
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 4
OF 9
5.(b)(1)
Contents - Type, form, and quantity of material (Continued) i.
the BPRAs have cooled for a minimum of 25 years, and ii.
the maximum exposure of the BPRA(s) shall be 30,000 Megawatt-Days per Metric Ton of Uranium (MWd/MTU).
VI.
Fuel may include thimble plug assemblies (TPAs) provided:
i.
the minimum cooling time of the TPAs is 25 years, ii.
the maximum exposure of the TPA(s) shall not exceed 125,000 MWd/MTU, and iii. only TPAs that do not have water displacement rods extending into the active fuel may be loaded into the cask.
VII.
The combined weight of a fuel assembly and any BPRA or TPA shall not exceed 1330 lbs.
VIII.
The combined weight of all fuel assemblies, BPRAs, and TPAs in a single cask shall not exceed 52,000 lbs.
IX.
The fuel shall not be a Damaged or Oxidized Fuel Assembly; a Damaged or Oxidized Fuel Assembly is:
a partial fuel assembly from which fuel pins are missing unless dummy fuel pins are used to displace an amount of water equal to or greater than that displaced by the original pins; has known or is suspected to have gross cladding failures (other than pinhole leaks) or have structural defects sufficiently severe to adversely affect fuel handling and transfer capability; or has been exposed to air oxidation during storage, as indicated by maintenance or operating records.
X.
The number of assemblies in the container shall not exceed 40.
XI.
The assembly average burnup shall be greater than or equal to the burnup calculated according to the following equations:
B = -1,259.8X2 + 20,242X - 23,617; for fuel assemblies with BPRA insertions during depletion B = -366.95X2 + 14,770X -17,200; for fuel assemblies without BPRA insertions during depletion Where:
B = Burnup (MWd/MTU),
X = Initial enrichment (weight percent (wt%) U-235)
XII.
The minimum cooling time for the fuel assemblies is 30 years. Content may include BPRAs or TPAs, which have a minimum cooling time of 25 years. Various combinations of minimum
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 5
OF 9
5.(b)(1) Contents - Type, form and quantity of material (Continued) assembly average enrichment and maximum assembly average burnup prior to transport shall be in accordance with Table 1-2 in this certificate.
XIII.
The maximum decay heat per fuel assembly shall not be more than 0.475 kW and 19 kW per package including the BPRAs and TPAs.
XIV.
The boron-10 (B-10) in the Boral neutron poison plates in the basket must be uniformly distributed in the plates with a minimum areal density of 10 mg/cm2.
XV.
Integral Fuel Burnable Absorber is not an authorized content.
XVI.
Fuel assemblies with the following irradiation history shall be authorized for transport:
i.
The minimum average specific power shall be 14 MW/Assembly, ii.
The minimum hot leg average moderator density shall be 0.705 g/cm3, iii.
The maximum hot leg average moderator temperature shall be 584 K (592°F),
iv.
The average fuel temperature shall not exceed 901 K (1,162°F), and v.
The maximum average soluble boron concentration shall not exceed 675 parts per million based on an average over the limiting non-linear boron letdown curve.
XVII.
The nominal length of the assembly axial blankets shall not exceed 6.2 in.
XVIII.
The maximum cooling time of the spent fuel shall not exceed 200 years.
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 6
OF 9
Table 1-1 Fuel Assembly Specifications1 Fuel Assembly Type Fuel Characteristics Exxon 14x14 Standard Exxon 14x14 High Burnup Exxon 14x14 TOPROD WE 14x14 Standard WE 14x14 OFA Max. Active Fuel Length (in.)
144 144 144 144 144 Max. Number of Fuel Rods per Assembly 179 179 179 179 179 Max. Fuel Rod Pitch (in.)
0.556 0.556 0.556 0.556 0.556 Min. Clad Thickness (in.)
0.0300 0.0310 0.0295 0.0243 0.0243 Min. Clad Outer Diameter (OD) (in.)
0.424 0.417 0.426 0.422 0.400 Clad Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Max. Pellet OD (in.)
0.3565 0.3565 0.3505 0.3659 0.3444 Min. Guide/Instrument Tube OD (in.)
16@0.541 1@0.424 16@0.541 1@0.424 16@0.541 1@0.424 16@0.539 1@0.422 16@0.528 1@0.4015 Max.
Guide/Instrument Tube Inner Diameter (in.)
16@0.507 1@0.374 16@0.507 1@0.374 16@0.507 1@0.374 16@0.505 1@0.3734 16@0.490 1@0.3499 Max. Assembly and BPRA Length (in.)
161.3 161.3 161.3 161.3 161.3 Max. Assembly Width (in.)
7.763 7.763 7.763 7.763 7.763 Maximum MTU/Assembly 0.380 0.380 0.380 0.410 0.380 Maximum Initial Assembly Average Enrichment (wt% U-235) 3.85 3.85 3.85 3.85 3.85 Maximum Assembly Average Burnup (MWd/MTU) 45,000 (see Table 1-22) 45,000 (see Table 1-2) 45,000 (see Table 1-2) 45,000 (see Table 1-2) 45,000 (see Table 1-2)
Minimum Cooling Time (years) 30 (see Table 1-2) 30 (see Table 1-2) 30 (see Table 1-2) 30 (see Table 1-2) 30 (see Table 1-2) 1.
Pre-irradiated nominal dimensions used in the design analyses and may be verified against as-built records.
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 7
OF 9
Table 1-2 Required Minimum Cooling Time for Spent Fuel Assemblies1,2,3,4 Minimum Assembly Average Initial Enrichment (wt.% U-235)
Maximum Assembly Average Burnup (GWd/MTU) 2 2.25 2.35 2.75 3
3.25 3.4 3.6 3.85 17 30 30 30 30 30 30 30 30 30 18 30 30 30 30 30 30 30 30 30 19 30 30 30 30 30 30 30 30 30 20 30 30 30 30 30 30 30 30 30 21 30 30 30 30 30 30 30 30 30 22 30 30 30 30 30 30 30 30 30 23 30 30 30 30 30 30 30 30 30 24 30 30 30 30 30 30 30 30 30 25 30 30 30 30 30 30 30 30 30 26 30 30 30 30 30 30 30 30 30 27 30 30 30 30 30 30 30 30 30 28 30 30 30 30 30 30 30 30 30 29 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 31 30 30 30 30 30 30 30 32 30 30 30 30 30 30 30 33 30 30 30 30 30 30 30 34 30 30 30 30 30 30 30 35 30 30 30 30 30 30 30 36 30 30 30 30 30 30 30 37 30 30 30 30 30 30 30 38 30 30 30 30 30 30 30 39 30 30 30 30 30 30 30 40 30 30 30 30 30 30 30 41 30 30 30 30 30 30 30 42 30 30 30 30 30 30 30 43 30 30 30 30 30 44 30 30 30 30 45 30 30 30 30
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 8
OF 9
Notes:
1.
For fuel characteristics that fall between the assembly average enrichment values in Table 1-2 of this certificate, use the next lower enrichment, and next higher burnup to determine minimum fuel cooling time.
2.
Fuel assemblies that were located in the Rod Cluster Control Assembly control bank D position during Unit 1 cycle 1 and Unit 2 cycle 1 shall have a minimum cooling time of greater than 35 years.
3.
The assembly average enrichment and the assembly average burnup are the enrichment and burnup averaged over the fuel assembly, including the axial blankets.
4.
Fuel assemblies with a maximum average burnup and a minimum average enrichment for which no cooling time is specified in the table are not authorized contents.
5.(c)
Criticality Safety Index:
0.0 6.
In addition to the requirements of Subpart G of 10 CFR Part 71:
(a)
The package must be prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the application, as supplemented.
(b)
Each packaging must be acceptance tested and maintained in accordance with the Acceptance Tests and Maintenance Program in Chapter 8 of the application, as supplemented.
(c)
The package contents shall be limited to the contents that were in storage in the package under SNM License No. 2506 (10 CFR Part 72) as of May 2011. Any additional reuse of the packaging after post-shipment unloading of the original content is prohibited.
(d)
This certificate applies to only the 29 TN-40 packages already fabricated and in use at the PINGP under SNM License No. 2506 (10 CFR Part 72).
(e)
Within 60 days of the first shipment of a shipping campaign involving any TN-40 package, the Certificate holder will notify the NRC of the leakage test method chosen to demonstrate compliance with the regulations in 10 CFR Part 71 related to leakage from the TN-40 package.
7.
Transport by air is not authorized.
8.
Packagings must be marked with Package Identification Number USA/9313/B(U)F-96.
9 The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 71.17.
- 10. The personnel barrier shall be installed at all times while transporting a loaded overpack.
- 11. Expiration date: December 31, 2028.
U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (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 9313 4
71-9313 USA/9313/B(U)F-96 9
OF 9
REFERENCES TN-40 Transportation Packaging Safety Analysis Report, Revision 17, November 2023 FOR THE U.S. NUCLEAR REGULATORY COMMISSION Yoira K. Diaz-Sanabria, Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Date: December 18, 2023 Signed by Diaz-Sanabria, Yoira on 12/18/23