ML18304A029
| ML18304A029 | |
| Person / Time | |
|---|---|
| Site: | 07109226 |
| Issue date: | 10/30/2018 |
| From: | John Mckirgan Spent Fuel Licensing Branch |
| To: | General Atomics |
| White B | |
| Shared Package | |
| ML18304A028 | List: |
| References | |
| EPID L-2018-RNW-0026 | |
| Download: ML18304A029 (9) | |
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 9226 5 71-9226 USA/9226/B(U)F-85 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 General Atomics General Atomics application dated August 31, 1994, 3550 General Atomics Court as supplemented San Diego, California 92121-1122
- 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.: GA-4 (2) Description The GA-4 Legal Weight Truck Spent Fuel Shipping Cask consists of the packaging (cask and impact limiters) and the radioactive contents. The packaging is designed to transport up to four intact pressurized-water reactor (PWR) irradiated spent fuel assemblies as authorized contents. The packaging includes the cask assembly and two impact limiters, each of which is attached to the cask with eight bolts. The overall dimensions of the packaging are approximately 90 inches in diameter and 234 inches long.
The containment system includes the cask body (cask body wall, flange, and bottom plate);
cask closure; closure bolts; gas sample valve body; drain valve; and primary O-ring seals for the closure, gas sample valve, and drain valve.
Cask Assembly The cask assembly includes the cask, the closure, and the closure bolts. Fuel spacers are also provided when shipping specified short fuel assemblies to limit the movement of the fuel. The cask is constructed of stainless steel, depleted uranium, and a hydrogenous neutron shield. The cask external dimensions are approximately 188 inches long and 40 inches in diameter. A fixed fuel support structure divides the cask cavity into four spent fuel compartments, each approximately 8.8 inches square and 167 inches long. The closure is recessed into the cask body and is attached to the cask flange with 12 1-inch diameter bolts.
The closure is approximately 26 inches square, 11 inches thick, and weighs about 1510 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 9226 5 71-9226 USA/9226/B(U)F-85 2 OF 9 5.a. (2) (continued)
The cask has two ports allowing access to the cask cavity. The closure lid has an integral half-inch diameter port (hereafter referred to as the gas sample valve) for gas sampling, venting, pressurizing, vacuum drying, leakage testing, or inerting. A 1-inch diameter port in the bottom plate allows draining, leakage testing, or filling the cavity with water. A separate drain valve opens and closes the port. The primary seals for the gas sample valve and drain valve are recessed from the outside cask surface as protection from punctures. The gas sample valve and the drain valve also have covers to protect them during transport.
Cask The cask includes the containment (flange, cask body, bottom plate and drain valve seals);
the cavity liner and fuel support structure; the impact limiter support structure; the trunnions and redundant lift sockets; the depleted uranium gamma shield; and the neutron shield and its outer shell. The cask body is square, with rounded corners and a transition to a round outer shell for the neutron shield. The cask has approximately a 1.5 inch thick stainless steel body wall, 2.6 inch thick depleted uranium shield (reduced at the corners), and 0.4 inch thick stainless steel fuel cavity liner.
The cruciform fuel support structure consists of stainless steel panels with boron-carbide (B4C) pellets for criticality control. A continuous series of holes in each panel, at right angles with the fuel support structure axis, provides cavities for the B4C pellets. The fuel support structure is welded to the cavity liner and is approximately 18 inches square by 166 inches long and weighs about 750 lbs.
The flange connects the cask body wall and fuel cavity liner at the top of the cask, and the bottom plate connects them at the bottom. The gamma shield is made up of five rings, which are assembled with zero axial tolerance clearance within the depleted uranium cavity, to minimize gaps. The impact limiter support structure is a slightly tapered 0.4 inch thick shell on each end of the cask. The shell mates with the impact limiters cavity and is connected to the cask body by 36 ribs.
The neutron shield is located between the cask body and the outer shell. The neutron shield design maintains continuous shielding immediately adjacent to the cask body under normal conditions of transport. The details of the design are proprietary. The design, in conjunction with the operating procedures, ensures the availability of the neutron shield to perform its function under normal conditions of transport.
Two lifting and tie-down trunnions are located about 34 inches from the top of the cask body, and another pair is located about the same distance from the bottom. The trunnion outside diameter is 10 inches, increasing to 11.5 inches at the cask interface. Two redundant lift sockets are located about 26 inches from the top of the cask body and are flush with the outer skin.
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 9226 5 71-9226 USA/9226/B(U)F-85 3 OF 9 5.a. (2) (continued)
Materials All major cask components are stainless steel, except the neutron shield, the depleted uranium gamma shield, and the B4C pellets contained in the fuel support structure. All O-ring seals are fabricated of ethylene propylene.
Impact Limiters The impact limiters are fabricated of aluminum honeycomb, completely enclosed by an all-welded austenitic stainless steel skin. Each of the two identical impact limiters is attached to the cask with eight bolts. Each impact limiter weighs approximately 2,000 lbs.
(3) Drawings The packaging is constructed and assembled in accordance with the following GA Drawing Number:
Drawing No. 031348, sheets 1 through 19, Revision D (Proprietary Version); GA-4 Spent Fuel Shipping Cask Packaging Assembly 5.(b) Contents (1) Type and Form of Material:
(a) Intact fuel assemblies. Fuel with known or suspected cladding defects greater than hairline cracks or pinhole leaks is not authorized for shipment.
(b) The fuel authorized for shipment in the GA-4 package is irradiated 14x14 and 15x15 PWR fuel assemblies with uranium oxide fuel pellets. Before irradiation, the maximum enrichment of any assembly to be transported is 3.15 percent by weight of uranium-235 ( 235U). The total initial uranium content is not to exceed 407 Kg per assembly for 14x14 arrays and 469 Kg per assembly for 15x15 arrays.
c) Fuel assemblies are authorized to be transported with or without control rods or other non-fuel assembly hardware (NFAH). Spacers shall be used for the specific fuel types, as shown on sheet 17 of the Drawings.
(d) The maximum burnup for each fuel assembly is 35,000 MWd/MTU with a minimum cooling time of 10 years and a minimum enrichment of 3.0 percent by weight of 235U or 45,000 MWd/MTU with a minimum cooling time of 15 years (no minimum enrichment).
(e) The maximum assembly decay heat of an individual assembly is 0.617 kW. The maximum total allowable cask heat load is 2.468 kW (including control components and other NFAH when present).
(f) The PWR fuel assembly types authorized for transport are listed in Table 1. All parameters are design nominal values.
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 9226 5 71-9226 USA/9226/B(U)F-85 4 OF 9 (2) Maximum Quantity of Material per Package (a) For material described in 5.b.(1): four (4) PWR fuel assemblies.
(b) For material described in 5.b.(1): the maximum assembly weight (including control components or other NFAH when present) is 1,662 lbs. The maximum weight of the cask contents (including control components or other NFAH when present) is 6,648 lbs., and the maximum gross weight of the package is 55,000 lbs.
Table 1 - PWR Fuel Assembly Characteristics Fuel Type Design No. of Fuel Rod Pellet Zr Clad Active Fuel Mfr.-Array Initial U Fuel Pitch (in.) Diameter Thickness Length (Versions) (kg/assy.) Rods (in.) (in.) (in.)
W-15x15 469 204 0.563 0.3659 0.0242 144 (Std/ZC)
W-15x15 463 204 0.563 0.3659 0.0242 144 (OFA)
BW-15x15 464 208 0.568 0.3686 0.0265 142 (Mk.B,BZ,BGD)
Exx/A-15x15 432 204 0.563 0.3565 0.030 144 (WE)
CE-15x15 413 204 0.550 0.358 0.026 144 (Palisades)
CE-14x14 376 176 0.580 0.3765 0.028 128 (Ft.Calhoun)
W-14x14 397 176 0.580 0.3805 0.026 137 (Model C)
CE-14x14 386 176 0.580 0.3765 0.028 137 (Std/Gen.)
Exx/A-14x14 381 176 0.580 0.370 0.031 137 (CE)
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 9226 5 71-9226 USA/9226/B(U)F-85 5 OF 9 5.b(2)(b)(continued)
Fuel Type Design No. of Fuel Rod Pellet Zr Clad Active Fuel Mfr.-Array Initial U Fuel Pitch (in.) Diameter Thickness Length (Versions) (kg/assy.) Rods (in.) (in.) (in.)
W-14x14 358 179 0.556 0.3444 0.0243 144 (OFA)
W-14x14 407 179 0.556 0.3674 0.0225 145.5 (Std/ZCA,/ZCB)
Exx/A-14x14 379 179 0.556 0.3505 0.030 142 (WE) 5.c. Criticality Safety Index (CSI): 100
- 6. Fuel assemblies with missing fuel pins shall not be shipped unless dummy fuel pins that displace an equal amount of water have been installed in the fuel assembly.
- 7. In addition to the requirements of Subpart G of 10 CFR 71:
- a. Each package shall be both prepared for shipment and operated in accordance with detailed written operating procedures. Procedures for both preparation and operation shall be developed using the specifications contained within the application. At a minimum, those procedures shall require the following provisions:
(1) Identification of the fuel to be loaded and independent verification that the fuel meets the specifications of Condition 5.b of the CoC.
(2) That before shipment the licensee shall:
(a) Perform a measured radiation survey to assure compliance with 49 CFR 173.441 and 10 CFR 71.47 and assure that the neutron and gamma measurement instruments are calibrated for the energy spectrums being emitted from the package.
(b) Verify that measured dose rates meet the following correlation to demonstrate compliance with the design bases calculated hypothetical accident dose rates:
3.4 x (peak neutron dose rate at any point on cask surface at its midlength) +
1.0 x (gamma dose rate at that location) 1000 mR/hr.
(c) Verify that the surface removable contamination levels meet the requirements of 49 CFR 173.443 and 10 CFR 71.87.
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 9226 5 71-9226 USA/9226/B(U)F-85 6 OF 9 7.a.(2) (continued)
(d) Inspect all containment seals and closure sealing surfaces for damage. Leak test all containment seals with a gas pressure rise test after final closure of the package. The leak test shall have a test sensitivity of at least 1 x10-3 standard cubic centimeters per second of air (std-cm3/sec) and there shall be no detectable pressure rise. A higher sensitivity acceptance and maintenance test may be required as discussed in Condition 7.b.(5), below.
(3) Before leak testing, the following closure bolt and valve torque specifications:
(a) The cask lid bolts shall be torqued to 235 +/- 15 ft-lbs.
(b) The gas sample valve and drain valve shall be torqued to 20 +/- 2 ft-lbs.
(4) During wet loading operations and prior to leak testing, the removal of water and residual moisture from the containment vessel in accordance with the following specifications:
(a) Cask evacuation to a pressure of 0.2 psia (10 mm Hg) or less for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
(b) Verifying that the cask pressure rise is less than 0.1 psi in 10 minutes.
(5) Before shipment, independent verification of the material condition of the neutron shield as described in SAR Section 7.1.1.4 or 7.1.2.4.
- b. All fabrication acceptance tests and maintenance shall be performed in accordance with detailed written procedures. Procedures for fabrication, acceptance testing, and maintenance shall be developed using the specifications contained within the application and shall include the following provisions:
(1) All containment boundary welds, except the final fabrication weld joint connecting the cask body wall to the bottom plate, shall be radiographed and liquid-penetrant examined in accordance with ASME Code Section III, Division 1, Subsection NB. Examination of the final fabrication weld joint connecting the cask body wall to the bottom plate may be ultrasonic and progressive liquid penetrant examined in lieu of radiographic and liquid penetrant examination.
(2) The upper lifting trunnions and redundant lifting sockets shall be load tested, in the cask axial direction, to 300 percent of their maximum working load (79,500 lbs. minimum) per trunnion and per lifting socket, in accordance with the requirements of ANSI N14.6. The upper and lower lifting trunnions shall be load tested, in the cask transverse direction, to 150 percent of their maximum working load (20,625 lbs. minimum) per trunnion, in accordance with the requirements of ANSI N14.6.
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 9226 5 71-9226 USA/9226/B(U)F-85 7 OF 9 7.b.(continued)
(3) The cask containment boundary shall be pressure tested to 1.5 times the Maximum Normal Operating Pressure of 80 psig. The minimum test pressure shall be 120 psig.
(4) All containment seals shall be replaced within the 12-month period prior to each shipment.
(5) A fabrication leakage test shall be performed on all containment components including the O-ring seals prior to first use. Additionally, all containment seals shall be leak tested after the third use of each package and within the 12-month period prior to each shipment. Any replaced or repaired containment system component shall be leak tested. The leakage tests shall verify that the containment boundary leakage rate does not exceed the design leakage rate of 1 x10-7 std-cm3/sec. The leak tests shall have a test sensitivity of at least 5 x 10-8 std-cm3/sec.
(6) The depleted uranium shield shall be gamma scanned with 100 percent inspection coverage during fabrication to ensure that there are no shielding discontinuities. The neutron shield supplier shall certify that the shield material meets the minimum specified requirements (proprietary) used in the applicants shielding analysis.
(7) Qualification and verification tests to demonstrate the crush strength of each aluminum honeycomb type and lot to be utilized in the impact limiters shall be performed.
(8) The boron carbide pellets, fuel support structure and fuel cavity dimensions, and 235U content in the depleted uranium shall be fabricated and verified to be within the specifications of Table 2 to ensure criticality safety.
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 9226 5 71-9226 USA/9226/B(U)F-85 8 OF 9 7.b.(8) (continued)
Table 2 Specified Parameter Minimum Maximum B4C boron enrichment 96 wt% 10B N/A Diameter of each B4C pellet 0.426 in 0.430 in Height of each B4C pellet stack 7.986 in 8.046 in Mass of 10B in each B4C pellet stack 31.5 g N/A Mass of each B4C pellet stack 43.0 g 45.0 g Diameter of each fuel support 0.432 in 0.44 in structure hole Fuel support structure nominal hole pitch N/A 0.55 in Fuel support structure hole depth minus B4C pellet-stack height 0.009 in 0.129 in (at room temperature)
Thickness of each fuel support 0.600 in 0.620 in structure panel Fuel cavity width N/A 9.135 in 235U content in depleted uranium N/A 0.2 wt%
shielding material
- 8. Transport of fissile material by air 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. Expiration Date: October 31, 2023.
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 9226 5 71-9226 USA/9226/B(U)F-85 9 OF 9 REFERENCES General Atomics Safety Analysis Report for the GA-4 Legal Weight Truck Spent Fuel Shipping Cask, August 31, 1994.
Supplements dated: August 5, 1998, General Atomics Safety Analysis Report for the GA-4 Legal Weight Truck Spent Fuel Shipping Cask, Revision G (Proprietary) and Revision H (Non-Proprietary), June 12, 2003; and September 24, 2008; October 25, 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 10/30/18