ML18330A248
| ML18330A248 | |
| Person / Time | |
|---|---|
| Site: | 07109255 |
| Issue date: | 11/28/2018 |
| From: | John Mckirgan Spent Fuel Licensing Branch |
| To: | Mathues G TN Americas LLC |
| Terry T | |
| Shared Package | |
| ML18330A247 | List: |
| References | |
| EPID L-2018-RNW-0025 | |
| Download: ML18330A248 (8) | |
Text
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 1 OF 8
- 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 Transnuclear, Inc. consolidated application dated 7135 Minstrel Way August 4, 2003, as supplemented.
Columbia, MD 21045
- 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.: NUHOMS MP187 Multi-Purpose Cask (2) Description The NUHOMS MP187 Multi-Purpose Cask (package) consists of an outer cask, into which one of the four different dry shielded canisters (DSC) is placed. During shipment, energy-absorbing impact limiters are utilized for additional package protection.
Cask The purpose of the cask is to provide containment and shielding of the radioactive materials contained within the DSC during shipment. The cask is constructed of stainless steel and lead with a neutron shield of cementitious material. The inside cavity of the cask is a nominal 68 inches in diameter and 187 inches long. The bottom access closure is approximately 5 inches thick and 17 inches in diameter, secured by 12 1-inch diameter bolts. The top closure is approximately 6.5 inches thick and is secured by 36 2-inch diameter bolts. Both closures are sealed by redundant O-rings.
Containment is provided by a stainless steel closure lid bolted to the stainless steel cask. The containment system of the NUHOMS MP187 transportation cask consists of (a) the inner shell, (b) the bottom end closure plate, (c) the top closure plate, (d) the top closure inner O-ring seal, (e) the ram closure plate, (f) the ram closure inner O-ring seal, (g) the vent port screw, (h) the vent port O-ring seal, (i) the drain port screw, and (j) the drain port O-ring seal. No credit is given to the DSC as a containment boundary.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 2 OF 8 Shielding is provided by 4 inches of stainless steel, 4 inches of lead, and approximately 4.3 inches of neutron shielding. The overall length of the cask is approximately 200 inches; the outer diameter is approximately 93 inches. The maximum gross weight of the package, with impact limiters, is approximately 282,000 lbs. The total length of the package with the impact limiters attached is approximately 308 inches. Four removable trunnions (two upper and two lower) are provided for handling and lifting.
Dry Shielded Canisters (DSCs)
The purpose of the DSC, which is placed within the transport cask, is to permit the transfer of spent fuel assemblies, into or out of a storage module, a dry transfer facility, or a pool as a unit. The DSC also provides additional axial biological shielding during handling and transport. The DSC consists of a stainless steel shell and a basket assembly. The approximately 5/8-inch thick shell has an outside diameter of about 67 inches and an external length of about 186 inches. The DSC basket assembly provides criticality control and contains a storage position for each fuel assembly. The basket is composed of circular spacer discs machined from thick carbon steel plates. Axial support for the DSC basket is provided by four high strength steel support rod assemblies. Carbon steel components of each DSC basket assembly are electrolytically coated with a thin layer of nickel to inhibit corrosion.
On the bottom of each DSC is a grapple ring, which is used to transfer a DSC horizontally from the cask into and out of dry storage modules. Because of the nature of the fuel that is to be transported, four different types of DSCs are designed for the package. Variations in the DSC configurations are summarized below:
- Fuel-Only Dry Shielded Canisters (FO-DSC)
The FO-DSC has a cavity length of approximately 167 inches and has solid carbon steel shield plugs at each end. The FO-DSC is designed to contain up to 24 intact Babcock and Wilcox (B&W) pressurized water reactor (PWR) spent fuel assemblies. The FO-DSC basket assembly consists of 24 guide sleeve assemblies with integral borated neutron absorbing plates, 26 spacer discs, and 4 support rod assemblies.
- Fuel/Control Components Dry Shielded Canister (FC-DSC)
The FC-DSC has an internal cavity length of approximately 173 inches to accommodate fuel with the B&W control components installed. To obtain the increased cavity length, the shield plugs are fabricated from a composite of lead and steel. The FC basket is similar to the FO-DSC except that the support rod assemblies and guide sleeves are approximately 6-inches longer. The FC-DSC is also designed to contain up to 24 intact B&W PWR spent fuel assemblies with control components.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 3 OF 8
- Failed Fuel Dry Shielded Canister (FF-DSC)
The FF-DSC has an internal cavity length of approximately 173 inches to accommodate 13 damaged B&W PWR spent fuel assemblies. Because the cladding has been locally degraded, individual (screened) fuel cans are provided to confine any gross loose material, maintain the geometry for criticality control, and facilitate loading and unloading operations. The FF-DSC is similar to FC-DSC in most respects with the exception of the basket assembly. The FF-DSC basket may be fabricated from austenitic stainless steel.
- 24PT1 Dry Shielded Canister (24PT1-DSC)
The 24PT1-DSC has an internal cavity length of approximately 167 inches with a solid carbon steel shield plug at each end. The 24PT1-DSC will accommodate 22 to 24 Westinghouse (WE) 14 x14 PWR spent fuel assemblies, including control components. Control components authorized that are integral to WE 14x14 fuel assemblies include rod cluster control assemblies, thimble plug assemblies, and neutron source assemblies only. Fuel assemblies may be damaged or intact as described in 5.b(2)(a). The 24PT1-DSC basket assembly consists of 24 guide sleeve assemblies with integral borated neutron absorbing plates, 26 spacer discs, and 4 support rod assemblies. Up to four screened individual failed fuel cans are provided for storage of damaged fuel within the guide sleeve assemblies. These failed fuel cans are similar in configuration to the FF-DSC failed fuel cans.
Impact Limiters The impact limiter shells are fabricated from stainless steel. Within that shell are closed-cell polyurethane foam and aluminum honeycomb material. The impact limiter is attached to the cask by carbon steel bolts. Each impact limiter is bolted to the cask body through the neutron shield top and bottom support rings. The weight of each impact limiter is approximately 15,800 lbs.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 4 OF 8 (3) Drawings The package shall be constructed and assembled in accordance with the following Transnuclear West Drawing Numbers:
NUH-05-4000NP, Revision 9, NUH-05-4004, Revision 16, Sheets 1 through 2 Sheets 1 through 5 MP187 Multi-Purpose Cask NUHOMS FO-DSC & FC-DSC General Arrangement PWR Fuel Main Assembly NUH-05-4001, Revision 15, NUH-05-4005, Revision 14, Sheets 1 through 6 Sheets 1 through 5 MP187 Multi-Purpose Cask NUHOMS FF-DSC Main Assembly PWR Fuel Main Assembly NUH-05-4002, Revision 5 NUH-05-4006NP, Revision 7, Sheets 1 and 2 Sheets 1 and 2 MP187 Multi-Purpose Cask NUHOMS MP187 Multi-Purpose Impact Limiters Transportation Skid/Personnel Barrier NH-05-4003, Revision 10, NUH-05-4010, Revision 2, Sheets 1 and 2 Sheets 1 through 6 NUHOMS MP187 Multi-Purpose NUHOMS - 24PT1-DSC Cask Main Assembly On-Site Transfer Arrangement (b) Contents of Packaging (1) Type and Form of Material (a) Intact fuel assemblies Intact fuel assemblies - Assemblies containing fuel rods with no known or suspected cladding defects greater than hairline cracks or pinhole leaks are authorized when contained in the FO-DSC, FC-DSC, or 24PT1-DSC.
(b) Damaged fuel assemblies - Assemblies containing fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks or with cracked, bulging, or discolored cladding are authorized when contained in a failed fuel can in the FF-DSC or the 24PT1-DSC. Spent fuel, with plutonium in excess of 20 curies per package, in the form of debris, particles, loose pellets, and fragmented rods or assemblies are not authorized. Damaged fuel assemblies may be shipped with or without control components.
(c) (i) The fuel authorized for shipment in the NUHOMS-MP187 FO, FC, or FF DSC is B&W 15x15 uranium oxide PWR fuel assemblies with a maximum initial pellet enrichment of 3.43% by weight of U235, and a total uranium content not to exceed 466 Kg per assembly.
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 5 OF 8 (ii) The fuel authorized for shipment in the NUHOMS-MP187 24PT1-DSC is WE 14x14 stainless steel clad (SC) or zircaloy clad mixed oxide (MOX) PWR fuel assemblies as described in Table 2.
(d) Intact B&W 15x15 fuel assemblies without control components shall be shipped only in the FO-DSC. Intact B&W 15x15 fuel assemblies with control components shall be shipped only in the FC-DSC.
(e) Intact WE 14x14 fuel assemblies with or without control components shall be shipped only in the 24PT1-DSC. Control components authorized are integral to WE 14x14 fuel assemblies include rod cluster control assemblies, thimble plug assemblies, and neutron source assemblies only.
(f) (i) The maximum burn-up and minimum cooling times for the individual B&W 15x15 assemblies shall meet the requirements of Table 1. In addition, the fuel shall have been decayed for a time sufficient to meet the thermal criteria of 5.b(1)(g) and (h).
The maximum total allowable cask heat load is 13.5 kW.
(ii) The maximum enrichment, burn-up and minimum cooling times for the individual WE 14x14 fuel assemblies shall meet the requirements of Table 2. In addition, the fuel shall have been decayed for a time sufficient to meet the thermal criteria of 5.b.(1)(g) and (h). The maximum total allowable cask heat load for the 24 PT1-DSC is per Table 2.
(g) (i) The maximum assembly decay heat (including control components when present) of B&W 15x15 individual fuel assembly is 0.764 kW, referred to as Type I, or 0.563 kW, referred to as Type II.
(ii) The maximum assembly decay heat (including control components when present) of WE 14x14 individual fuel assembly is per Table 2.
(h) (i) Control components for B&W 15x15 fuel assemblies stored in the FO, FC and FF-DSCs shall be cooled for at least 8 years.
(ii) Control components for WE 14x14 fuel assemblies stored in the 24PT1-DSC shall be cooled for at least 10 years.
(2) Maximum quantity of material per package (a) (i) For material described in 5.b(1) to be stored in the FO, FC or FF-DSCs: 24 PWR intact fuel assemblies or 13 damaged fuel assemblies, with no more than 15 damaged fuel rods per assembly. Where a DSC is to be loaded with fewer fuel assemblies than the DSC capacity, dummy fuel assemblies with the same nominal weight as a standard fuel assembly shall be installed in the unoccupied spaces.
(ii) For material described in 5.b(1) to be stored in the 24PT1-DSC: 22 to 24 PWR fuel assemblies of which up to four may be damaged WE 14x14 SC fuel assemblies with the balance intact WE 14x14 SC or MOX fuel assemblies. No more than one
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 6 OF 8 damaged WE 14x14 MOX fuel assembly can be stored per 24PT1-DSC with the balance intact WE 14x14 SC fuel assemblies. The damaged fuel assemblies shall have no more than 14 damaged fuel rods per assembly and shall be stored in the four outer corner fuel assembly locations along the 45°, 135°, 225°, 315° azimuth of the 24PT1-DSC. A DSC may include two empty slots if they are located on symmetrically opposite locations with respect to the 0° - 180° and 90°-270° DSC axes.
Any additional empty fuel slots shall be loaded with dummy fuel assemblies that displace the same or greater amount of volume and with the same nominal weight as a standard fuel assembly. Fuel spacers shall be located at the bottom and top of each fuel assembly to center the fuel assemblies within the DSC. Failed fuel cans require only bottom spacers since a top spacer is integral to each failed fuel can.
(b) For material described in 5.b(1): the approximate maximum payload (including control components when present) is 81,100 lbs.
Table 1- FO, FC and FF-DSC Fuel Assembly Burn-up vs. Cooling Time Maximum Minimum Minimum Minimum Maximum Minimum Minimum Minimum Burn-up Enrichment Required Required Burn-up Enrichment Required Required (MWD/MTIHM)* in the Type I Type II (MWD/MTIHM)* in the Active Type I Type II Active Fuel Cooling Cooling Fuel Region Cooling Cooling Region Time Time (w/o U-235) Time Time (w/o U-235) (years) (years) (years) (years)
<23,200 n/a 5 5 33,000 2.90 7 10 23,200 2.38 5 5 34,000 2.95 7 11 24,000 2.43 5 6 35,000 2.67 7 14 25,000 2.49 5 6 35,000 2.99 7 11 26,000 2.55 5 7 36,000 3.03 8 13 27,000 2.61 5 7 37,000 3.00 8 14 28,000 2.66 5 8 37,000 3.07 8 14 29,000 2.00 6 10 38,000 3.11 9 15 29,000 2.71 5 8 39,000 3.15 9 16 30,000 2.76 5 8 40,000 3.19 9 17 31,000 2.81 6 9 32,000 2.86 6 10
- Megawatt Days per Metric Ton of Initial Heavy Metal
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 7 OF 8 Table 2 - 24PT1-DSC Fuel Assembly Burnup vs. Cooling Time Minimum Cooling Time / Max Maximum Minimum Maximum Heat Load Per Cask / Max Fuel Type Enrichment Enrichment Burnup Assembly Heat Load (Weight %) (Weight %) (MWD/ MTU)
(Incl. Control Components1)
WE 14x14 Stainless Steel 45,000 Clad (SC) 3.76 235U (May include Integral Fuel 38 years/14 kW/
4.05 235U 40,000 Burnable Absorber, boron 3.36 235U 0.583 kW coated fuel pellets) 35,000 3.12 235U 0.71 235U 2.84 fissile Pu (64 2.78 fissile Pu rods) (64 rods) 30 years/13.706 kW/
WE 14x14 MOX 3.10 fissile Pu (92 3.05 fissile Pu 25,000 0.294 kW rods) (92 rods) 3.31 fissile Pu (24 3.25 fissile Pu rods) (24 rods)
Notes:
1 Control component cooling time must be a minimum of 10 years.
(c) Criticality Safety Index 0
- 6. Type I fuel assemblies shall be loaded only into the four innermost cells of a DSC, while Type II assemblies may be loaded into any cell when using the FO-DSC or the FC-DSC. The FF-DSC has no Type I or II placement restrictions. The 24PT1-DSC has restrictions on the location of damaged fuel assemblies per Section 5.b.(2).
- 7. For operating controls and procedures, in addition to the requirements of Subpart G of 10 CFR Part 71:
(a) Each package shall be both prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the application, as supplemented.
(b) All fabrication acceptance tests and maintenance shall be performed in accordance with the Acceptance Tests and Maintenance Program in Chapter 8 , as supplemented. In addition, this shall include:
(1) With the exception of the weld between the inner shell and top forging, all longitudinal and circumferential inner shell welds, which form the containment boundary of the cask, shall be radiographically inspected (RT) with acceptance standards in accordance with the ASME Code,Section III, Division 1, NB-5320. The weld between the inner shell and top forging shall be verified by RT or ultrasonically inspected (UT). The substitution of UT for the examination of the completed weld may be made provided the examination is performed using detailed written procedures, proven by actual demonstration to the satisfaction of the inspector as capable of detecting and locating defects described in ASME Code,Section III, Division 1 Subsection NB (2) Verification of the DSC outer top cover plate weld by either volumetric or multilayer penetrant test (PT) examination. If PT is used, at a minimum, it must include the root, each
NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9255 14 71-9255 USA/9255/B(U)F-85 8 OF 8 successive 1/4 inch weld thickness, and the final layer. The inspection of the weld must be performed by qualified personnel and shall meet the acceptance requirements of ASME B&PVC Section III, NB-5350. The inspection process, including findings (indications) shall be made a permanent part of the licensees records by video, photographic, or other means providing an equivalent retrievable record of weld integrity.
(3) The minimum lead thickness in the main cask body, away from the trunnions and the top and bottom forgings, shall be 3.90 inches.
(4) The neutron shield shall have a minimum thickness of 4.31 inches.
- 8. This package is approved for exclusive use by rail, truck, or marine transport. Transport by air of fissile material is not authorized.
- 9. The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 71.17.
- 10. Fabrication of new packagings is not authorized.
- 11. Revision No. 13 of this certificate may be used until November 30, 2019.
- 12. Expiration Date: November 30, 2023.
REFERENCES Transnuclear, Inc. application dated August 4, 2003.
Supplement(s) dated September 16, 2008; July 26, 2013; January 27, 2014; November 18, 2016, and October 10, 2018.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
/RA/
John McKirgan, Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Date: 11/28/18