ML18311A020

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Enclosure 2: Safety Evaluation Report (Letter to W. Fowler Revision 19 of Certificate of Compliance No. 9235 for the Model No. NAC-STC Package)
ML18311A020
Person / Time
Site: 07109235
Issue date: 11/07/2018
From: John Mckirgan
Spent Fuel Licensing Branch
To: Fowler W
NAC International
White B
Shared Package
ML18311A016 List:
References
EPID L-2017-LLA-0066, EPID L-2017-LLA-0418
Download: ML18311A020 (26)
v • d • e


Text

SAFETY EVALUATION REPORT Docket No. 71-9235 Model No. NAC-STC Certificate of Compliance No. 71-9235 Revision 19 Summary By applications dated March 16, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17079A512) and December 8, 2017 (ADAMS Accession No. ML17353A026), as supplemented on July 17, 2017 (ADAMS Accession No. ML17200C956), September 20, 2017 (ADAMS Accession No. ML17265A143), March 6, 2018 (ADAMS Accession No. ML18067A133), June 5, 2018 (ADAMS Accession No. ML18158A192),

June 21, 2018 (ADAMS Accession No. ML18176A099), July 18, 2018 (ADAMS Accession No. ML18201A204), August 21, 2018 (ADAMS Accession No. ML18235A503), September 19, 2018 (ADAMS Accession No. ML18264A231), October 12, 2018 (ADAMS Accession No. ML18289A398), NAC International, Inc., and NAC email on October 26, 2018 (ADAMS Accession No. ML18299A096), (NAC or the applicant) requested revision to Certificate of Compliance No. 9235, for the Model No. NAC-STC package.

In addition to editorial changes to ensure consistency with the new information, NAC revised 19 drawings, submitted one new drawing, and requested changes to the safety analysis report (SAR) referenced in the certificate of compliance, as described below:

  • Revise operating procedures in SAR Section 7 to accommodate the shield ring addition;
  • Revise SAR Section 8.1.3 to permit leak testing without the inner lid and inner lid vent and drain port cover plates installed and added a requirement to perform leakage testing of the inner lid and with the inner lid vent and drain port cover plates during the final fabrication leakage tests;
  • Revise SAR Section 8.1.4.3 to allow several acceptable methods of verifying weld integrity of the impact limiters;
  • Revise SAR Section 8.1.5.1 to generalize the gamma scan detector size and spacing requirements and authorize ultrasonic testing of the packages inner closure lid, inner bottom forging, outer closure lid and cask outer bottom plate to demonstrate their gamma shielding effectiveness;
  • Revise SAR Table 8.2-1 to include the post-fabrication thermal performance test and radial neutron shield shell visual inspections;
  • Revise SAR Section 8.2.6 by deleting the periodic thermal performance test and replacing it with a post-fabrication thermal performance test that shall be done in accordance with the fabrication acceptance test method prior to subsequent use if the cask experiences certain conditions;
  • Revise SAR Section 8.4.1 to modify the longitudinal seam weld offset requirements and allow the use of any calibrated measuring and test equipment (M&TE in the SAR) when inspecting the diameter and cylindricity of the inner shell bore; Enclosure 2
  • Correct cool down rate unit conversions in SAR Section 8.4.2; and
  • Revise Section 8.4.3 to include an alternate lead pour procedure.

The staff used the guidance in NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, as well as associated ISG documents to perform the review of the proposed packaging changes. Based on the statements and representations in the application, as supplemented, and the conditions listed in the following chapters, the staff concludes that the package meets the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.

EVALUATION 1.0 GENERAL INFORMATION 1.1 Packaging Description The packaging body is made of two concentric stainless steel shells. The inner shell is 1.5 inches thick and has an inside diameter of 71 inches. The outer shell is 2.65 inches thick and has an outside diameter of 86.7 inches. The annulus between the inner and outer shells is filled with lead.

The inner and outer shells are welded to steel forgings at the top and bottom ends of the packaging. The bottom end of the packaging consists of two stainless steel circular plates which are welded to the bottom end forging. The inner bottom plate is 6.2 inches thick and the outer bottom plate is 5.45 inches thick. The space between the two bottom plates is filled with a 2-inch thick disk of a synthetic polymer (NS4FR) neutron shielding material.

The packaging is closed by two steel lids which are bolted to the upper end forging. The inner lid (containment boundary) is 9 inches thick and is made of Type 304 stainless steel. The outer lid is 5.25 inches thick and is made of SA-705 Type 630, H1150 (17-4PH) stainless steel. The inner lid is fastened by forty two, 1-1/2-inch diameter bolts and the outer lid is fastened by thirty six, 1-inch diameter bolts. The inner lid is sealed by two O-ring seals. The outer lid is equipped with a single O-ring seal. The inner lid is fitted with a vent and drain port which are sealed by O-rings and cover plates. The containment system seals may be metallic or Viton. Viton seals are used only for directly-loaded fuel that is to be shipped without subsequent long-term interim storage.

The packaging body is surrounded by a 1/4-inch thick jacket shell constructed of 24 stainless steel plates. The jacket shell is 99 inches in diameter and is supported by 24 longitudinal stainless steel fins which are connected to the outer shell of the packaging body. Copper plates are bonded to the fins. The space between the fins is filled with NS4FR shielding material.

Four lifting trunnions are welded to the top end forging. The package is shipped in a horizontal orientation and is supported by a cradle under the top forging and by two trunnion sockets located near the bottom end of the packaging.

The package is equipped at each end with an impact limiter made of redwood and balsa. Two impact limiter designs consisting of a combination of redwood and balsa wood, encased in Type 304 stainless steel, are provided to limit the g-loads acting on the package during an accident. The predominantly balsa wood impact limiter is designed for use with all the proposed contents. The predominately redwood impact limiters may only be used with directly loaded fuel or the Yankee-multi-purpose canister (MPC) configuration.

The package includes a stainless steel ring assembly which, when applicable, includes a top shield ring and shear ring. The stainless steel ring assembly is installed on the upper cask body, between the top impact limiter and the neutron shield shell in the upper region of the packaging. The shield ring consists of four sectors: bottom sector, top sector and two side sectors. The bottom sector of the shield ring assembly is an SA-705, Type 630, 17-4PH stainless steel forging. The top sector and side sectors are fabricated from SA-240, Type 304 stainless steel. The bolt material is SA-193, Grade B6, Type 410 stainless steel for all bolts.

1.2 Contents The applicant requested changes to the cooling times for directly-loaded uncanistered high burnup (HBU) and low burnup (LBU) fuel to be transported with the shield ring, but did not request changes to the maximum decay heat. The contents consist of undamaged 17x17 Advanced Fuel Assembly pressurized-water reactor (PWR) LBU (i.e., assembly average burnup less than or equal to 45 GWd/MTU) and undamaged 17x17 Advanced Fuel Assembly PWR HBU (i.e., assembly average burnup exceeding 45 GWd/MTU) fuel assemblies that meet the fuel assembly criteria for Framatome-Cogema 17x17 fuel listed in Table 1 in the certificate of compliance.

For the LBU assemblies, the maximum assembly heat load may not exceed 850 watts, and the maximum assembly-average burnup may not exceed 45 GWd/MTU, provided the loading pattern meets the requirements of Configuration A, B or C, as shown in NAC International Drawing No. 423-800. The minimum fuel assembly cool time is determined from Table 1. The use of the shield ring assembly, as configured in NAC International Drawing No. 423-927, is required.

For the HBU assemblies, the maximum assembly heat load may not exceed 1.71 kW, and the maximum assembly-average burnup may not exceed 55 GWd/MTU, provided the loading pattern meets the requirements of Configuration A, B or C, as shown in NAC International Drawing No. 423-800. The minimum fuel assembly cool time is determined from Tables 2 through 4, of this SER. The use of the shield ring assembly, as configured in NAC International Drawing No. 423-927, is required for the new LBU cool times given in Table 1. Only zircaloy-4 and M5 zirconium-based alloy cladding may be loaded per shipment, with a maximum of 4 zircaloy-4 fuel assemblies per shipment. Gadolinium based integral fuel burnable absorber rods (IFBAs) are permitted, but boron-based IFBAs are not. The maximum time duration from the time the package breaks the surface of the spent fuel pool until the package is placed in the horizontal orientation is limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If this time limit cannot be met, the package may be re-flooded. HBU fuel assemblies subjected to a package re-flood are not authorized for shipment. HBU fuel shipments are limited to a total duration of 6 months from the time package loading is complete until the package arrives at its final destination. These time limits also apply to packages containing commingled loadings of HBU fuel and LBU fuel. The minimum fuel assembly cool time is determined from Tables 2 through 4, depending on loading configuration.

The fuel assemblies shall not have been previously stored in an independent spent fuel storage installation licensed under 10 CFR Part 72, Licensing Requirements for the Independent

Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

The new fuel cooling tables for the HBU fuel assemblies are shown in Tables 1 through 4, below (listed as Tables 6 through 9 in the CoC).

Table 1 - Fuel Cool Time Table (17x17 PWR LBU)

Minimum Fuel Cool Time in Years (Cobalt content 1.2 g/kg)

Min. initial Assembly Average Burnup [GWd/MTU]

Assembly B10 10<B15 15<B20 20<B25 25<B30 30<B 32.5<B 35<B 37.5<B 40<B41 41<B42 42<B43 43<B44 44<B45 Avg. Enr. [wt.

32.5 35 37.5 40

%]

1.7 E < 1.9 4.0 4.0 4.0 4.5 5.9 7.2 9.8 1.9 E < 2.1 4.0 4.0 4.0 4.4 5.5 6.4 8.3 11.4 15.3 2.1 E < 2.3 4.0 4.0 4.0 4.3 5.2 5.9 7.2 9.7 13.2 2.3 E < 2.5 4.0 4.0 4.0 4.2 4.9 5.6 6.6 8.4 11.4 12.8 14.3 15.9 17.6 19.2 2.5 E < 2.7 4.0 4.0 4.0 4.1 4.8 5.3 6.0 7.4 9.8 11.1 12.5 13.9 15.5 17.1 2.7 E < 2.9 4.0 4.0 4.0 4.0 4.7 5.0 5.7 6.7 8.5 9.6 10.8 12.1 13.6 15.1 2.9 E < 3.1 4.0 4.0 4.0 4.0 4.6 5.0 5.6 6.2 7.6 8.4 9.4 10.6 11.9 13.3 3.1 E < 3.3 4.0 4.0 4.0 4.0 4.6 5.0 5.5 6.0 6.9 7.6 8.3 9.2 10.4 11.7 3.3 E < 3.5 4.0 4.0 4.0 4.0 4.6 4.9 5.4 6.0 6.7 7.0 7.5 8.2 9.1 10.2 3.5 E < 3.7 4.0 4.0 4.0 4.0 4.5 4.9 5.4 5.9 6.6 6.9 7.2 7.6 8.2 9.0 3.7 E < 3.9 4.0 4.0 4.0 4.0 4.5 4.9 5.3 5.9 6.5 6.8 7.1 7.5 7.9 8.4 3.9 E < 4.1 4.0 4.0 4.0 4.0 4.5 4.8 5.3 5.8 6.5 6.8 7.0 7.4 7.8 8.3 4.1 E < 4.3 4.0 4.0 4.0 4.0 4.5 4.8 5.3 5.8 6.4 6.7 7.0 7.4 7.7 8.1 4.3 E < 4.5 4.0 4.0 4.0 4.0 4.4 4.8 5.2 5.8 6.4 6.6 6.9 7.3 7.7 8.1

Table 2 - Revised Fuel Cool Time Table (Configuration A 17x17 PWR HBU)

Minimum Fuel Cool Time in Years (Cobalt content 1.2 g/kg)

Min. initial Assembly Average Burnup [GWd/MTU]

Assembly Avg.

Enr. [wt. %] 45<B46 46<B47 47<B48 48<B49 49<B50 50<B51 51<B52 52<B53 53<B54 54<B55 2.9 E < 3.1 4.0 4.0 4.5 5.0 5.7 6.3 6.9 7.6 8.4 -

3.1 E < 3.3 4.0 4.0 4.0 4.3 4.8 5.4 6.0 6.7 7.4 8.1 3.3 E < 3.5 4.0 4.0 4.0 4.1 4.2 4.7 5.2 5.8 6.4 7.1 3.5 E < 3.7 4.0 4.0 4.0 4.0 4.1 4.2 4.5 5.0 5.6 6.2 3.7 E < 3.9 4.0 4.0 4.0 4.0 4.1 4.2 4.3 4.4 4.8 5.4 3.9 E < 4.1 4.0 4.0 4.0 4.0 4.0 4.1 4.2 4.3 4.5 4.7 4.1 E < 4.3 4.0 4.0 4.0 4.0 4.0 4.1 4.2 4.3 4.4 4.5 4.3 E 4.5 4.0 4.0 4.0 4.0 4.0 4.0 4.1 4.2 4.4 4.5 Table 3 - Revised Fuel Cool Time Table (Configuration B 17x17 PWR HBU)

Minimum Fuel Cool Time in Years (Cobalt content 1.2 g/kg)

Min. initial Assembly Average Burnup [GWd/MTU]

Assembly Avg. Enr. 45<B46 46<B47 47<B48 48<B49 49<B50 50<B51 51<B52 52<B53 53<B54 54<B55

[wt. %]

2.9 E < 3.1 4.4 4.9 5.5 6.1 6.8 7.6 8.3 9.1 10.0 -

3.1 E < 3.3 4.4 4.5 4.7 5.3 5.9 6.6 7.3 8.0 8.8 9.7 3.3 E < 3.5 4.3 4.4 4.5 4.7 5.1 5.7 6.3 7.0 7.8 8.6 3.5 E < 3.7 4.2 4.4 4.5 4.6 4.7 4.9 5.5 6.1 6.8 7.5 3.7 E < 3.9 4.2 4.3 4.4 4.5 4.7 4.8 4.9 5.3 5.9 6.6 3.9 E < 4.1 4.1 4.3 4.4 4.5 4.6 4.8 4.9 5.0 5.2 5.7 4.1 E < 4.3 4.1 4.2 4.3 4.4 4.5 4.7 4.8 5.0 5.1 5.3 4.3 E 4.5 4.0 4.2 4.3 4.4 4.5 4.6 4.8 4.9 5.0 5.2

Table 4 - Revised Fuel Cool Time Table (Configuration C 17x17 PWR HBU)

Minimum Fuel Cool Time in Years (Cobalt content 1.2 g/kg)

Min. initial Assembly Average Burnup [GWd/MTU]

Assembly Avg.

45<B46 46<B47 47<B48 48<B49 49<B50 50<B51 51<B52 52<B53 53<B54 54<B55 Enr. [wt. %]

2.9 E < 3.1 7.4 8.2 9.1 10.0 11.0 12.0 13.1 14.3 15.5 -

3.1 E < 3.3 6.4 7.1 7.9 8.8 9.7 10.7 11.6 12.7 13.9 15.1 3.3 E < 3.5 5.5 6.2 6.9 7.7 8.5 9.4 10.4 11.3 12.4 13.5 3.5 E < 3.7 5.4 5.6 6.0 6.7 7.5 8.3 9.2 10.1 11.0 12.0 3.7 E < 3.9 5.3 5.5 5.7 5.9 6.6 7.3 8.1 8.9 9.8 10.8 3.9 E < 4.1 5.2 5.4 5.6 5.8 6.0 6.4 7.1 7.9 8.7 9.6 4.1 E < 4.3 5.2 5.4 5.6 5.7 5.9 6.1 6.4 7 7.7 8.5 4.3 E 4.5 5.1 5.3 5.5 5.7 5.9 6.0 6.3 6.6 6.8 7.6

1.3 Drawings NAC International revised 19 drawings and submitted one new drawing showing the approved transport configurations for the proposed changes.

The revised drawings showing the transport packaging include:

423-800, Sheets 1-3, Rev. 20P and 20NP Cask Assembly - NAC-STC Cask 423-802, Sheets 1-7, Rev. 26 Cask Body - NAC-STC Cask 423-803, Sheets 1-2, Rev. 15 Lid Assembly - Inner, NAC-STC Cask 423-804, Sheets 1-2, Rev. 12 Details - Inner Lid, NAC-STC Cask 423-805, Sheets 1-2, Rev. 9 Details - Inner Lid, NAC-STC Cask 423-806, Sheets 1-2, Rev. 14 Port Coverplate Assy - Inner Lid, NAC-STC Cask 423-807, Sheets 1-3, Rev. 6 Assembly, Port Cover, NAC-STC Cask 423-811, Sheets 1-2, Rev. 13 Details -NAC-STC Cask 423-812, Rev. 7 Nameplates - NAC-STC Cask 423-859, Rev. 1 Attachment Hardware, Balsa Limiters, NAC-STC 423-870, Rev. 8 Fuel Basket Assembly, PWR, 26 Element, NAC-STC Cask 423-874, Rev. 3 Heat Transfer Disk, Fuel Basket, PWR, 26 Element, NAC-STC Cask 423-878, Sheets 1-2, Rev. 5 Alternate Tube Assembly, NAC-STC Cask 423-880, Rev. 3P Shielded Thermal Shunt Assembly, NAC-STC Cask 423-900, Rev. 9 Package Assembly Transportation, NAC-STC Cask 423-209, Rev. 2 Impact Limiter Assy - Upper, NAC-STC Cask 423-210, Rev. 2 Impact Limiter Assy-Lower, NAC-STC Cask 423-257, Rev. 3 Balsa Impact Limiter, Upper, NAC-STC Cask and 423-258, Rev. 3 Balsa Impact Limiter, Lower, NAC-STC Cask.

New drawing:

423-927, Rev. 1P & 2NP Shield Ring Assembly, NAC-STC Cask 2.0 STRUCTURAL EVALUATION The objective of the structural review is to verify that the structural performance of the package has been adequately evaluated and meets the requirements of 10 CFR Part 71, including the tests and conditions for normal conditions of transport and hypothetical accident conditions.

2.1 Alternate Lead Pour Procedure In support of the alternate lead pour procedure in Section 8.4.3, NAC submitted revised calculations of the stresses induced by the increased lead temperature from 750 °F (in standard lead pour procedure) to 790 °F (in the alternate lead pour procedure).

The applicant provided calculations for thermal stresses in the SAR Chapter 2, Section 2.6.11 "Fabrication Conditions" including the thermal stresses to the inner and outer shells before, during, and after the lead pour evolution. The applicant used the same equations for thermal stress calculations that were previously reviewed and accepted by the NRC staff.

Enclosure 2

The NRC staff reviewed the new calculations and found that the change in thermal stresses due to the temperature change were small and negligible when compared with the overall magnitude of the thermal stresses in the package. The staff concluded that the changes in thermal stress demonstrated in the application do not affect the capability of the package to meet the structural requirements of 10 CFR Part 71.

2.2 Impact Limiter Changes NAC submitted Drawing Nos. 423-209, Rev. No. 2 and 423-210, Rev. No. 2 for the redwood impact limiters to add tolerances for the screw tube circle diameter and the impact limiter shell section. The NRC staff reviewed the Drawing Nos. 423-209, Rev. No. 2 and 423-210, Rev. No. 2 and noted that the requested tolerances were +0.20/-0.02 inch for each dimension (the smallest of which is 44 inches) and concluded that the tolerances additions are acceptable because they do not affect the capability of the package to meet the structural requirements of 10 CFR Part 71.

2.3 Shield Ring NAC added a shield ring to the top forging which is intended to reduce the dose rates at the cask surface during the transport of directly-loaded HBU and LBU fuel to meet the normal condition external surface dose rate limit of 200 mrem/hr applicable to an exclusive use open transport configuration without crediting the personnel barrier. The scope of this structural review is limited to the evaluation of the NAC-STC shield ring under normal conditions of transport, specifically, the 1-foot free drop analysis of 10 CFR 71.71(c)(7), since the shield ring is not needed to meet the dose rate requirements after hypothetical accident conditions.

The applicant stated that the shield ring consists of a bottom sector, a top sector and two side sectors. The bottom sector is an SA-705, Type 630, 17-4PH stainless steel forging. The top and side sectors are fabricated from SA-240, Type 304 stainless steel. The top and bottom sectors are connected to the cask top forging with eight, 1-8 UNC bolts. The side sector sub-components are connected to the top and bottom sectors with four 1/2-13 UNC bolts.

2.3.2 Codes and Standards The applicant used the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NF, to qualify the bolts and related components.

2.3.3 Design Criteria The applicant used the allowable stress criteria of ASME B&PV Code,Section III, Division 1, Subsection NF, and the material properties of the bolts and shield ring at 250 °F (121 °C) to evaluate the structural adequacy of the components.

2.3.4 Loads and Load Combinations The applicant evaluated the connections using the 20g deceleration values for the 1-foot side and end free drops from Section 2.6.7.4, Table 2.6.7.4.1-3 of the SAR.

2.3.5 Analytical Approach The applicant used hand calculations to determine the stress in the bolts as a result of the inertial loads applied to the components from the 1-foot side and end drop evaluations. For tensile, shear and bearing stress evaluations, the applicant determined a factor of safety by dividing the allowable stress by the calculated stress. If the factor of safety is greater than 1.0, then the structural performance is acceptable as shown.

= 1.0 The applicant also considered shear and tensile stress acting together by calculating an interaction ratio in accordance with ASME B&PV Code. As long as the interaction ratio (shown below) is less than or equal to 1.0, the structural performance of the bolt is acceptable.

= + 1.0 2.4 Normal Conditions of Transport: Free Drop The applicant evaluated the shield ring for the free drop for normal conditions of transport. The applicant stated that the bolts are only loaded in shear as a result of the end drop. The applicant also calculated the bearing stress on the bolts as a result of the end drop. The applicant reported minimum factors of safety of 2.53 and 4.5 for shear and bearing stresses respectively in the bolts.

For the side drop evaluation, the applicant considered three different orientations.

Orientation #1 aligns the impact region with one of the 1-8 UNC bolts. Orientation #2 aligns the impact region between two of the 1-8 UNC bolts. Orientation #3 aligns the impact region with two of the four 1/2-13 UNC bolts that attach the side sector plate to the top and bottom plate.

In Section 2.6.7.8 of the SAR, the applicant reports the factors of safety for 1-8 UNC bolted connection and the 1/2-13 UNC bolted connection for orientations #1 and #3, as well as the interaction ratios. The applicant considered bolt shear stress, bolt external thread shear stress, shield ring top and bottom sector internal thread shear, bearing stress on the shield ring sector under the bolt head, and shear stress in the shield ring side sector due to the bolt head. All factors of safety were greater than 1.0. Because the maximum load on the 1-8 UNC bolts for orientation #2 is less than the maximum load on the bolts for orientation #2, the applicant only determined the stress interaction ratio for orientation #1. The applicant reported stress interaction ratios of less than 1.0 for all three orientations for shear and tension.

2.5 Materials The applicant corrected the mechanical properties for forged SA-336 type 304 stainless steel listed in SAR Table 2.3.2-2. The applicant also corrected the referenced tables in the ASME B&PV Code for the mechanical properties for SA-240 type 304 stainless steel listed in SAR Table 2.3.2-1, forged SA-336 type 304 stainless steel listed in SAR Table 2.3.2-2, and Type XM-19 stainless steel listed in SAR Table 2.3.2-3.

The staff reviewed the mechanical properties provided in the corrected tables and confirmed that the values for ultimate strength, yield strength, design stress intensity, modulus of elasticity, coefficient of thermal expansion, and Poissons ratio are consistent with the values in the respective tables in ASME B&PV Code Section II Part D. The staff confirmed that the values for stress limits under alternating stresses is consistent with the values in the ASME B&PV Code Section III Appendix I. Staff confirmed that the material property values for all materials as a function of temperature, including the ultimate strength, yield strength, design stress intensity, modulus of elasticity, and coefficient of thermal expansion, were consistent with the values in the ASME B&PV Code Section II Part D. Staff confirmed that the values for Poisons ratio for all material were consistent with the values listed in ASME B&PV Code Section II Part D. Staff confirmed that the listed values for densities for all materials were accurate.

Additionally, the applicant evaluated the use of wood preservatives in SAR Section 2.4.4.2.5 and general requirements for lubrication and grease (provided the alternatives meet performance and compositional requirements of the nuclear power industry) in SAR Section 2.4.4.2.7. NAC stated that the wood preservatives are standard applications of preservatives and adhesives, so no post-application reactions will occur. Also, NAC stated that other lubricants and greases may be used as alternatives provided they meet the performance and general compositional requirements of the nuclear power industry, thus eliminating any potential reactions. The NRC staff reviewed the proposed changes and found that they are acceptable and will not result in significant chemical, galvanic or other reaction and will meet the requirements of 10 CFR 71.43(d).

2.5.1 Shield Ring The staff reviewed the new and revised design-basis drawings, including the proposed shield ring assembly drawing. The shield ring on the top forging reduces the dose rate at the cask surface when transporting directly-loaded HBU or LBU fuel with reduced cool times.

The shield ring assembly is manufactured from all stainless steel materials, which have been previously approved for use in the NAC-STC package. The at-temperature mechanical property values, as obtained from ASME B&PV Code, have been previously reviewed and approved in prior package amendments or in Section 2.5 of this SER (Tables 2.3.2-1, 2.3.3-1, 2.3.4-2 of the application). The staff confirmed that the bill of materials in the design-basis drawings and Section 2.3 of the application adequately define all construction materials, grades, and mechanical properties.

The staff verified that the pertinent drawing identified the acceptance criteria and non-destructive examination for all safety-related welding for the shield ring assembly. The applicant clarified that the identified weld connects the lift lug to the top sector of the shield ring and is used for handling the top sector for installation and removal. The weld and tab are not credited for structural or shielding performance under transport conditions. Accordingly, the details of the weld non-destructive examination are deemed nonessential details for the licensing drawing, as they do not relate to the safety performance of the transport package.

The staff confirmed that the location and relative placement of the package attachment bolts and side-top/bottom sector bolts were properly identified in the shield ring drawing. The applicant revised the drawing to clarify the angular location of the package bolts and the shield rings counterbore depth, which determines the thread length that extends beyond the inner diameter of the ring sector as well as ensures the bolt head remains recessed.

The shield ring assembly is made from the same material (stainless steel) as the top forging and is machined to match the actual diameter of the cask top forging, which ensures that the shield ring is in contact with the cask top forging. The radiation effects on the stainless steel material for the top forging have been previously reviewed in prior approvals of the NAC-STC package.

The staff considers the conclusions on this discussion to bound the radiation effects on the shield ring assembly.

2.5.2 Ceramic Fiber Paper The applicant provided a revised fire-transient analysis for the directly-loaded fuel package configuration in order to demonstrate that the previously-approved ceramic fiber paper is optional (Calculation 423-3000, Revision 5). The ceramic fiber insulation serves to insulate the corners of the lead gamma shielding at each end of the package during the hypothetical fire accident scenario per the requirement in 10 CFR 71.73(c)(4). The staff confirmed that the optionality of the ceramic fiber insulation and insulation cover is properly reflected in the pertinent drawing. The staff notes that this amendment did not provide supporting thermal analyses to remove the ceramic fiber insulation and insulation cover for the transport of multipurpose canisters in the NAC-STC package.

The staff finds that the materials and materials specifications for the package comply with the requirement in 10 CFR 71.33(a)(5).

The staff reviewed the revised design-basis drawings and determined that the revisions do not result in credible mechanisms for chemical or galvanic reactions in the package during loading operations or during transport, which could compromise the intended functions of the structural materials. The stainless steel materials, previously approved for use in the package, have a long history of non-galvanic behavior within close proximity of each other when exposed to the normally-encountered environments during fuel loading and package transport operations. The staff also confirmed that the ceramic fiber material is heat-treated to remove volatile organics that could enhance adverse reactions.

The ceramic fiber paper used for lead insulation is temporarily fixed to the upper forging by a caulk material prior to attachment of a stainless steel sheath. The applicant clarified that the radiation resistance of the caulk material is not important to safety since it does not serve a purpose once the stainless steel sheathing is installed.

The staff finds that the package complies with the requirement in 10 CFR 71.43(d).

2.6 Corrections The applicant also proposed three minor changes in the revised SAR: (i) a correction of a typographical error in SAR Table 2.3.2-2, (ii) a correction of an error in Section 2.6.7.5.4.1, "Bolt Stress Evaluation," and (iii) removal of polytetrafluoroethylene (PTFE) O-rings, since they are no longer used on the NAC-STC package for containment.

The NRC staff reviewed the proposed changes and found that they are typographical or minor errors and are acceptable because the corrections requested by the applicant do not affect the capability of the package to meet the structural requirements of 10 CFR Part 71.

2.7 Findings

The staff has reviewed the package and concludes that the applicant has met the requirements of 10 CFR 71.33. The applicant described the materials used in the transportation package in sufficient detail to support the staffs evaluation.

The staff has reviewed the package and concludes that the applicant has met the requirements of 10 CFR 71.31(c). The applicant identified the applicable codes and standards for the design, fabrication, testing, and maintenance of the package and, in the absence of codes and standards, has adequately described controls for material qualification and fabrication.

The staff has reviewed the packages structural design description and concludes that the contents of the SAR meet the requirements of 10 CFR 71.31(a)(1) and (a)(2) as well as 10 CFR 71.33(a) and (b).

The staff reviewed the design criteria for the various components of the shield ring assembly and the structural codes and standards used in the package design and finds that they are acceptable because established codes and standards are identified in a manner that is consistent with NUREG-1617 and NUREG/CR-3854, Fabrication Criteria for Shipping Containers, and therefore meet the requirements of 10 CFR 71.31(c).

Based on review of the statements and representations in the application, the NRC concludes that the materials used in the transportation package design have been adequately described and evaluated and that the package meets the requirements of 10 CFR Part 71.

The staff reviewed the loads and load combinations and finds that they are acceptable as they are consistent with Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, and therefore meet the requirements of 10 CFR 71.71.

The staff reviewed the hand calculations as described in Section 2.4 of this SER, Normal Conditions of Transport: Free Drop, and determined the factor of safety and interaction ratio met the acceptance criteria, and therefore meets the requirements of 10 CFR 71.31.

Because the factors of safety for all bolts are greater than 1.0, and because the stress interaction ratios are less than 1.0, the staff determined that the structural performance of the bolted connections of the shield ring assembly are adequate for the 1-foot free drop, and concludes that there will be no substantial reduction in the effectiveness of the packaging that would prevent it from satisfying the requirements of 10 CFR 71.51(a)(1) for a Type B package.

The staff finds that the package has been evaluated to ensure there is no significant chemical, galvanic, or other reaction and complies with the requirement in 10 CFR 71.43(d).

Based on a review of the statements and representations in the application, the NRC staff concludes that the structural design has been adequately described and evaluated, and that the package has adequate structural integrity to meet the requirements of 10 CFR Part 71.

3.0 THERMAL EVALUATION The objective of the thermal review is to verify that the thermal performance of the package has been adequately evaluated for the tests specified under normal conditions of transport and

hypothetical accident conditions and that the package design satisfies the thermal requirements of 10 CFR Part 71. The staff reviewed the changes to the thermal design characteristics and thermal analyses for the NAC-STC package proposed by the amendment requests.

3.1 Neutron Shield Plates The amendment requested an alternate configuration with an increase of the thickness of the radial neutron shield stainless steel plate and copper plate as indicated on Drawing No.

423-802, Revision 25, Note 7. The applicant stated that thicker plates will increase the effective thermal conductivity for the radial neutron shield, thereby providing a slightly more effective path for heat rejection from the package shell to the ambient during normal conditions of transport.

The applicant concluded the maximum temperatures presented in SAR Tables 3.4-1, 3.4-2, and 3.4-3 for normal conditions of transport would be slightly reduced and therefore remain bounding. NRC staff agrees with NACs conclusion.

For the fire hypothetical accident conditions, the applicant stated that the heat input is maximized during the 30-minute fire condition due to the inclusion of the neutron shielding material solid synthetic polymer and considered voided at the end of the fire transient. Due to the thicker copper plates and stainless steel, the heat rejection after the fire is slightly enhanced.

As previously discussed, the normal conditions of transport temperatures that are used for the initial condition for the fire transient analysis are slightly reduced due to the higher effective conductivity of the radial neutron shield. The applicant concluded that the drawing change to allow for thicker stainless steel and copper plates would have an insignificant impact on the maximum component temperatures for the fire test in the hypothetical accident conditions presented in Table 3.5-1 of the SAR.

To demonstrate that the thicker plates would have a negligible impact, the applicant provided, in the response to the NRC request for additional information (RAI) 3.1, a proprietary sensitivity study using a 180° half-symmetry, full-length, three-dimensional thermal model with the NAC-STC HBU configuration that corresponded to the limiting configuration and decay heat.

The only change in the model is in the effective conductivity of the radial neutron shield due to the increased thickness of the stainless steel and copper plates. The applicants thermal analysis sensitivity study results, described in Appendix AA of NAC Calculation No. 423-3000, Rev. 6, shows that the change in radial neutron shield effective conductivity results in slight decreases of the maximum fuel temperatures during normal conditions of transport and hypothetical accident conditions.

The staff reviewed the applicants ANSYS thermal model and confirmed the increase in radial neutron shield thermal conductivity; therefore, the staff accepts that the maximum temperatures provided in Tables 3.4-1, 3.4-2, 3.4-3, and 3.5-1 of the SAR remain bounding for LBU fuel due to the conservatism in the calculation, and the sensitivity study for HBU fuel showed slight decreases in peak cladding temperature for normal conditions of transport and hypothetical accident conditions.

3.2 Shield Ring The application also included the following changes relevant to the thermal review: 1) including a shield ring on the top forging to reduce the dose rate to below 200 mrem/hr, 2) making the use of ceramic fiber paper at the corners of the lead shield, 3) updating O-ring location

temperatures, and 4) providing an alternative Viton O-ring material specification, Parker VM125-75.

The application described a stainless steel shield ring which may be bolted to the top forging of the cask and characterized the shield ring as insignificant on the thermal performance of the package. The applicant justified this by stating the shield ring is machined to match the diameter of the cask top forging to ensure contact, has a slightly larger surface area for transferring heat to the ambient by convection and radiation, and is thermally equivalent to increasing the diameter of the top forging.

3.3 Ceramic Fiber Paper The staff assessed the thermal impact on the Viton O-ring temperatures of making the ceramic fiber paper, Fiberfrax 972-H, optional for the directly loaded NAC-STC. The ceramic fiber paper provides insulation at the corners of the lead gamma shield to prevent it from reaching its melting point. The applicant provided a revised hypothetical accident conditions transient fire analysis. The initial conditions for the transient analysis was the steady-state normal conditions of transport initial conditions. The fire transient lasted for 30 minutes with a 1475 °F (800 °C) boundary condition, followed by a 64-hour cooldown. The staff reviewed the results of the analysis that showed without the presence of the ceramic fiber paper, the lead temperature is at 556 °F (291 °C) which is below the 600 °F (316 °C) limit specified by the applicant. Therefore, the staff concludes the lead remains within the allowable temperature limit.

The staff assessed the thermal impact of making the optional ceramic fiber paper on the Viton O-ring temperatures. The applicant post-processed the ANSYS fire cooldown analysis to use exact O-ring locations. The results showed that the port cover Viton O-ring maximum temperature is 312 °F (156 °C) and the inner lid Viton O-ring maximum temperature is 418 °F (214 °C) for less than 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using updated O-ring locations. The inner lid O-ring maximum temperature is above the steady state allowable temperature limit of 400 °F (204 °C), but is less than the manufacture provided test data for a 70-hour, dry heat resistance test at a temperature above 528 °F (276 °C) for Parker O-ring material specification VM835-75, and 482 °F (250 °C) for Parker O-ring material specification VM125-75.

The staff confirmed from the material specification data sheet that the Parker O-ring material specification VM125-75 minimum allowable service temperature is -40 °F (-40 °C). Therefore, the staff concludes that the containment boundary Viton O-ring for specification VM125-75 remains within the allowable temperature limit, and therefore is acceptable for use in this package.

3.4 Thermal Materials The ceramic fiber paper used for lead insulation is temporarily fixed to the upper forging by a caulk material prior to attachment of a stainless steel sheath. The applicant clarified that the thermal resistance of the caulk material is not important to safety since it does not serve a purpose once the stainless steel sheathing is installed.

The thermal properties for all other materials (i.e., stainless steel subcomponents of the shield ring assembly) were not revised from prior approvals of the NAC-STC package.

The staff confirmed that references for the technical specifications of package components were identified, as they pertain to the revisions in this amendment. Consistent with the guidance in

NUREG-1617, the staff confirmed that the minimum allowable service temperature of all components is less than or equal to -40 °F (-40 °C).

The applicant provided a revised fire-transient analysis for the directly-loaded fuel package configuration in order to demonstrate that the previously-approved ceramic fiber paper is optional (Calculation Package No. 423-3000, Revision 5). The ceramic fiber insulation serves to insulate the corners of the lead gamma shielding at each end of the cask during the hypothetical fire accident scenario per the requirement in 10 CFR 71.73(c)(4). The revised analysis yields higher temperatures for the inner lid inner O-ring and the portcover O-ring for the configuration without the ceramic fiber paper. The temperatures reported exceed the manufacturers recommended limit for a limited period of time. In order to support the acceptable performance of the O-rings, the applicant provided additional manufacturer test data including dry heat resistance properties at a temperature well above the calculated maximum inner lid inner O-ring temperature during the fire transient scenario. In addition, the applicant provided vendors seal life at-temperature data for the O-ring material, which supports adequate O-ring performance with a significant safety margin; therefore, the staff concludes that the O-ring seal performance is not compromised by the excursion to a temperature higher than the limit for the O-ring seal for a limited time.

The maximum allowable temperatures for all other component materials that could affect the containment, shielding, and criticality functions of the package from prior approvals of the package were not revised.

3.5 Findings

Based on a review of the statements and representations in the application, the staff concludes that changes to the thermal design proposed in the amendment, including material properties and component specifications, have been adequately described and evaluated, and that the performance of the package meets the thermal requirements of 10 CFR Part 71.

4.0 CONTAINMENT The objective of the containment review is to verify that the package design satisfies the containment requirements of 10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions. The staff reviewed the changes to the containment design characteristics and containment analyses for the NAC-STC, proposed by the amendment requests.

The application included the following changes relevant to the containment review: 1) making the use of ceramic fiber paper at the corners of the lead shield optional, 2) updating O-ring location temperatures, 3) providing an alternative Viton O-ring material specification, Parker VM125-75, and 4) removal of some of the references to PTFE O-rings within Chapters 1, 2, and 8 of the SAR.

The application described making the ceramic fiber paper, Fiberfrax 972-H, optional for the directly loaded NAC-STC package. The applicant provided hypothetical accident conditions transient fire analysis and the results showed that without the presence of the ceramic fiber paper, the containment boundary port cover Viton O-ring maximum temperature is 312 °F (156 °C) and the containment boundary inner lid Viton O-ring maximum temperature is 418 °F (214 °C) for less than 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using updated O-ring locations.

As described in Section 3.0 of this SER, the inner lid O-ring maximum temperature is above the steady state allowable temperature limit of 400 °F (204 °C), but is less than the manufacture provided test data for a 70-hour, dry heat resistance test at a temperature above 528 °F (276 °C) for VM825-75, and 482 °F (250 °C) for Parker O-ring material specification VM125-75.

The staff confirmed that the Viton O-ring material technical specification data sheets were provided in the SAR, and the Parker O-ring VM125-75 material compound was described on the licensing drawings. Therefore, the staff concludes the containment boundary Viton O-rings remain within temperature limits and their ability to perform their containment function is not impacted.

The staff confirmed that many references to PTFE O-rings were removed from Chapters 1, 2, and 8 of the SAR, PTFE O-rings do not appear on the licensing drawings, previously approved Viton O-rings are used on the licensing drawings, and the SAR remains consistent with the previous approval as a result of this change.

Based on a review of the statements and representations in the application, the staff concludes that the containment design has been adequately described and evaluated and that the package design meets the containment requirements of 10 CFR Part 71.

5.0 SHIELDING EVALUATION The objective of the review is to verify that the Model NAC-STC package provides adequate protection against direct radiation from its contents and that the package design meets the external radiation requirements of 10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions.

5.1 Ultrasonic Testing The applicant proposed to use ultrasonic (UT) for examination of the NAC-STC packaging system fabrication quality of the gamma shields, which include the inner closure lid, outer closure lid, inner bottom forging, and outer bottom plate, in place of the currently used gamma scan method as approved in Section 8.1.5.1, Gamma Shield Test of the SAR based on the guidance of NUREG/CR-3854.

The applicant revised the SAR to document this new method of fabrication test for the NAC-STC. The applicant also added in the drawings reference to the ASME B&PV code acceptance criteria for the gamma shield test and maintenance program.

5.2 Shield Ring The applicant requested to revise the CoC to add a shield ring on the top with the purpose of reducing the dose rate to below 200 mrem/hr, at the cask surface when shipping HBU or LBU fuel with reduced cool times. The shield ring assembly, as configured in NAC International Drawing No. 423-927, is required for specified contents and cooling times.

5.2.1 Source Specification According to the applicant, in order to generate a minimum cooling time table for HBU fuel, each fuel assembly is analyzed over a range of burnups, initial U-235 enrichments, and cooling times.

Fuel assembly burnup is evaluated from 45,000 MWd/MTU to 60,000 MWd/MTU in

1,000 MWd/MTU increments. Initial U-235 enrichments are evaluated from 2.9 to 4.9 wt %

U-235 in 0.2 wt % increments. Cooling times range from 4 to 60 years with varying increments.

The applicant performed source terms calculations for the design basis spent fuel for the NAC-STC HBU which is Westinghouse 17x17 PWR fuel, in the directly loaded basket of the NAC-STC with a shield ring. Also, partial loads as shown in configurations A, B, and C were evaluated. The PWR fuel types intended for transport in the directly loaded NAC-STC were analyzed at HBUs in order to determine the minimum cool times based on decay heat and dose rate limits. Cask total heat load is restricted to 24 kW.

The source terms for the fuel assemblies are calculated in SAS2H for burnups from 10 GWd/MTU to 60 GWd/MTU, enrichment between 1.0 to 4.9 wt % U-235, and cooling times between 0.5 to 90 years. These parameters are used to provide a full set of minimum cooling times for the shield/shear ring configuration. The NAC-STC models and fuel assembly source terms are imported from the previously approved NAC calculations. The models are modified to include the top shield ring. MCNP6, a transport theory-based Monte Carlo code, was employed by the applicant for this evaluation. The applicant developed dose rate response functions as a function of energy for each source region.

5.2.2 Geometry Considerations The applicant states that all model assumptions from the NAC-STC HBU shielding analysis previously used are retained for this evaluation. Additional assumptions are made for this evaluation such as the stainless steel shield ring assembly is simplified to a cylinder in the MCNP model and the neutron shield heat fins are modeled with an 18 mm thickness. According to the applicant, for the first assumption the shield ring ends at the upper trunnions and overlaps the shear ring. The use of the shield ring assembly, as configured in NAC International Drawing No. 423-927, is required for the new LBU cool times given in Table 1. HBU-only shipments with the new cool time tables do not require the shield ring as the package was evaluated with and without the shield ring installed. HBU and LBU commingled shipments require the shield ring when the LBU shipped utilizes the cooling times in Table 1. The upper trunnions are not included in the shielding model. The applicant states that any gaps between the shield ring and the upper trunnions or the shear ring are small and do not have a significant impact on calculated dose rates. For the second assumption, the dose rate increase with the thicker heat fins is not significant. These assumptions are based on dose rates calculations previously performed by the applicant.

5.2.3 Normal Conditions of Transport Also, to achieve the 200 mrem/hr limit at the surface of a flat-bed vehicle with cooling times shorter than those approved by the NRC in Revision 18 of Certificate of Compliance No. 9235, shielding evaluations were conducted to allow a reduced minimum cooling time when the shield ring configuration is used or to increase the cooling time and limit the fuel hardware cobalt content. Section 5.8 of the SAR contains evaluations of directly loaded HBU contents with and without the shield ring configuration, while Section 5.9 of the SAR contains the evaluation of directly loaded LBU contents with the shield ring configuration using cooling times lower than the currently approved. The HBU contents with the shield ring for configurations A, B, and C, under normal conditions of transport is determined to have a bounding dose rate of 190.0 mrem/hr radially at flat-bed vehicle surface.

The applicant imported loading tables from the previous revision which meet 10 CFR 71 dose rate limits. The loading tables were recalculated to evaluate the impact of the shield ring on the surface dose rates. The maximum dose rates on the surface and at 2m are calculated for all possible burnup, cooling time, and enrichment combinations by folding the response functions with the source spectra. The range of enrichments considered for each burnup is shown in Table 6-3 of the SAR. A minimum cooling time of 4 years is applied.

The staff evaluated the shielding design of the NAC STC package with shield ring and the dose rate tables produced for burnups from 10 GWd/MTU to 60 GWd/MTU. The 50 GWd/MTU loading tables are shown in Table 6-8 of the SAR for configuration A, Table 6-9 of the SAR for Configuration B, and Table 6-10 of the SAR for configuration C. The full loading tables are listed in Table 6-7 of the SAR. The associated cooling time tables are shown in Table 6-11 through Table 6-13 of the SAR. Consistent with NUREG-1617, NAC used the 1977 American National Standards Institute Standard Flux-To-Dose Rate Factors.

5.2.4 Hypothetical Accident Conditions For hypothetical accident conditions, the applicant considered gas depletion of the radial neutron shield, postulated redistribution of the cask lead shielding as a result of a drop accident, and removal of the impact limiters in the model and evaluated all new contents without the shield ring. The cask lead shield is assumed to redistribute as a result of the plastic deformation of the lead shielding and consequent filling of the narrow gap that forms between the lead and cask outer shell during fabrication. Both axial and radial slump conditions are analyzed. In the axial case, the lead shift results from a postulated cask end drop and causes an annular gap to appear in the lead shield at both ends of the cask. In the radial case resulting from a cask side drop, the potential maximum reduction at any single section of the lead shielding is modeled around the circumference of the shield. The staff determined that applying the maximum reduction at each end and over the circumference simultaneously is conservative for the accident condition shielding model.

The HBU contents without the shield ring for configuration A, loading under hypothetical accident conditions, is determined to have a dose rate of 704.1 mrem/hr at 1 meter and 9 years cool time. For configuration B loading, the maximum dose rate is 673.6 mrem/hr at 1 meter and 10 years cool time. For configuration C loading, the maximum dose rate is 823.1 mrem/hr at 1 meter and 10 years cool time.

5.3 Findings

The staff reviewed all the assumptions and parameters for the shield ring provided by the applicant. Also, the staff evaluated the loading tables that were generated with the shield ring to evaluate the impact of the shield ring on the surface dose rates. The staff found that the dose rates for HBU fuel when the shield ring is in placed still below the regulatory limit.

The method used by the applicant to calculate radiation source terms and dose rate is consistent with accepted industry practices and standards.

The staff reviewed the maximum dose rates for package under normal conditions of transport and hypothetical accident conditions and finds that the values are below the regulatory limit as prescribed in 10 CFR 71.47 and 71.51.

Based on its review of the application, the technical basis, and a comparison with the ASME B&PV code, as discussed in Chapter 8, below, and the recommendation of NUREG/CR-3854, the staff finds that using ultrasonic tests in place of gamma scan tests is an acceptable and can provide a reasonable assurance that the package design continue to meets the regulatory requirements of 10 CFR 71.43 and 10 CFR 71.51 on the dose rates of the radiations directly form the content of the package.

In addition, based on the review of applicants shielding safety analyses of the addition of the shield ring, the staff finds the change to be consistent with the appropriate codes and standards and NRC guidance, and that the package with the shield ring design satisfies the dose rate limits set forth in 10 CFR Part 71.47 and 71.51(a)(2).

6.0 CRITICALITY EVALUATION

There were no changes that affected the packages criticality evaluations.

7.0 PACKAGE OPERATIONS The applicant made changes in Chapter 7 of the application related to the installation and removal of the shield ring to ensure that the package will be prepared for shipment and operated in a manner consistent with the package design.

The primary changes to SAR Chapter 7 include ensuring proper installation and removal of the shield ring prior to and just after being loaded. The removal procedure involves the use of a lifting sling and then verifying the correct removal of lock wires, hex bolts, and socket head screws from the shield ring assembly.

NAC revised the preparation for transport procedures to include details of applying the shield ring after being loaded.

7.1 Findings

The staff confirmed that the operating procedures include appropriate language to verify that proper removal has taken place.

The staff reviewed and evaluated the revised loading procedures for directly-loaded, HBU spent fuel. Based on the statements and representations in the application, the staff concluded that the package operations meet the requirements of 10 CFR Part 71, and that they are adequate to assure the package will be operated in a manner consistent with its evaluation for approval.

Further, the certificate is conditioned to specify that the package must be prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the SAR, as supplemented.

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW Chapter 8 of the application identifies the acceptance tests and maintenance programs to be conducted on the Model No. NAC-STC package and verifies their compliance with the requirements of 10 CFR Part 71.

8.1 Leak Testing The applicant proposed to revise the leakage test requirements in SAR Section 8.1.3, Leakage Tests, to permit testing without the inner lid and inner lid vent and drain port coverplates installed. The staff verified that the applicant included a requirement to perform leakage testing of the inner lid and with the inner lid vent and drain port coverplates during the final fabrication leakage tests in Section 8.1.3.2, Final Fabrication Leakage Rate Testing, of the SAR.

The applicant also revised Section 8.1.4.2, Gaskets, of the SAR to describe that containment boundary Viton O-rings shall be replaced annually, or prior to transport if the containment boundary Viton O-rings have been installed longer than 1 year. The staff finds that this is consistent with NUREG-1617 that describes elastomeric seals should be replaced at an interval not to exceed one year, therefore the annual replacement time period in the SAR should not exceed 1 year.

8.2 Impact Limiter Component Tests The staff reviewed the revised methods for testing of the transport impact limiter stainless steel shell weld integrity, weld seam offset requirements, and specified measurement and test equipment to measure shell dimensions provided in SAR Section 8.1.4.3. The staff reviewed the testing requirements and acceptance criteria and determined that these changes are acceptable for verifying construction and maintenance of the NAC-STC transportation package impact limiter.

8.3 Ultrasonic Testing The applicant revised Section 8.1.5.1, Gamma Shield Test of the NAC-STC transportation package SAR. The applicant stated that demonstration for soundness as gamma shielding would be conducted using methods identified in NUREG/CR-3854, Section 3.2.1 which states that gamma scanning, or probing may be used to demonstrate the soundness of gamma shielding. NUREG/CR-3854, Section 3.2.1 also states that as an alternative to gamma scanning, ultrasonic testing of the gamma shielding materials may be used. The applicant stated that the cask body (i.e., inner shell and outer shell) containing lead shielding will be gamma scanned following the lead pour to demonstrate its soundness as gamma shielding. An equivalent mockup of the applicable cask body components will be made for calibration of equipment used to perform the gamma scanning.

The applicant stated that the package inner closure lid and inner bottom forging are obtained as ASME B&PV Code Section III, Division I, Subsection NB components and are examined using UT as part of procurement prior to final installation. The applicant stated that since the package inner closure lid and inner bottom forging are examined by UT per the requirements of ASME Section III, Division I, Subsection NB these components are acceptable gamma shields per NUREG/CR-3854 Section 3.2.1 without the need for being subjected to an additional gamma scan examination.

The applicant stated that, since the package outer closure lid, outer bottom forging, and outer bottom plate are not part of the containment boundary, they are not required by the ASME B&PV code to be examined by UT. The applicant stated that although not required, UT examination of these components will be conducted to demonstrate their gamma shielding effectiveness in accordance with the fabrication criteria identified in NUREG/CR-3854 Section 3.2.1. The applicant stated that the forged outer closure lid and the outer bottom forging will be examined by UT per ASME B&PV Section III NB-2542.1 using the acceptance standards of Section NB-2542.2. In addition, the applicant stated that the wrought package outer bottom plate will be examined by UT per ASME B&PV Section III NB-2532.1 using the acceptance standards of Section NB-2532.1(b). The applicant revised the NAC-STC SAR Section 8.1.5.1 to include the requirements to examine these components by UT.

The NRC staff reviewed the revisions to the gamma shield test in the acceptance tests for the NAC-STC transportation package SAR, the NAC-STC drawings, and the relevant sections of the ASME B&PV Code. The staff reviewed the NAC-STC system design and identified that both the package inner closure lid and inner bottom forging are produced from forged stainless steel. The staff also note that the examination of forgings and bars are required by ASME B&PV Code Section III NB-2540 to be ultrasonically examined prior to fabrication. The staff note that NB-2540 cites the requirement of ASME B&PV Code Section V Article 5 which in turn references several UT examination specifications included in ASME Section II for specific product forms. Heavy steel forgings are required to be inspected using ASME B&PV Code Section II, SA-388 using straight beam UT examination. Because of the varied nature of product forms that may be produced, ASME SA-388 does not include specific acceptance standards but does include guidance for development of acceptance standards. Acceptance standards are included in ASME B&PV Code Section III NB-2542.2. The staff note the straight beam UT method is not sufficient to determine the thickness or composition of the discontinuity, however, the thickness of any discontinuity in forged material will be thin relative to the component thickness based on the nature of the forming process. Therefore, the acceptance standards of NB-2542.2 are sufficient to demonstrate material integrity and adequacy for gamma shielding. The staff determined that the examination of the forged material by UT as required by ASME B&PV Code Section III NB-2540 is sufficient to demonstrate gamma shielding effectiveness without an additional gamma scan examination.

The staff reviewed the NAC-STC system design and identified that the package outer closure lid and outer bottom forging are produced from forged stainless steel. The staff determined that UT examination per the requirements of ASME B&PV Code Section III NB-2540 prior to fabrication is appropriate for these components because the thickness of any discontinuity in forged material will be thin relative to the plate component thickness based on the nature of the forming process. The staff determined that the revised fabrication specifications by the applicant to include examination of the forged outer closure lid and the outer bottom forging by UT per ASME B&PV Code Section III NB-2540 with the acceptance standards of ASME B&PV Code Section III NB-2542.2 is sufficient to demonstrate gamma shielding effectiveness of these components without an additional gamma scan examination.

The staff reviewed the NAC-STC system design and identified that the package outer bottom plate is produced from wrought stainless steel plate. In accordance with ASME B&PV Code Section III NB-2532.1, wrought plate materials greater than 2 inches thick are required to be inspected per ASME B&PV Code Section II, SA-578 using straight beam ultrasonic examination and meet the acceptance standards identified in NB-2532.1(b). This examination is intended to detect internal discontinuities parallel to the surface. The staff note the straight beam UT method is not sufficient to determine the thickness or composition of the discontinuity.

However, for rolled plate, the thickness of any discontinuity is expected to be thin relative to the plate thickness based on the nature of the forming process. The staff determined that the examination of the cask outer bottom plate material by UT as required by ASME B&PV Code Section III NB-2532.1 with the acceptance standards of NB-2532.1(b) is sufficient to demonstrate gamma shielding effectiveness without an additional gamma scan examination.

The staff reviewed the drawings for the NAC-STC package and note that the package outer bottom plate and package inner bottom forging are welded to the package outer bottom forging.

The welds are ultrasonically examined per ASME B&PV Code Section V Article 5 with acceptance per Section VIII Division 1, Article UW-53, which in turn referencesSection VIII Appendix 12. The applicant stated that per the acceptance criteria for the weld examination, any indications that are characterized as cracks, lack of fusion, or incomplete penetration are unacceptable regardless of length. Any voids in the weld that would impact shielding capability would be rejected based on the acceptance criteria. Therefore, the absence of voids within the weld joint and welds that meet the minimum weld size will provide adequate gamma shielding.

The staff reviewed the NAC-STC package body drawings and the referenced ASME B&PV Code sections. The staff also reviewed the acceptance criteria in ASME B&PV Code Section VIII Appendix 12 which states that indications characterized as cracks, lack of fusion, incomplete penetration are not acceptable. For other imperfections identified by UT, ASME Section VIII Appendix 12 includes evaluation requirements and acceptance criteria based on the size of the imperfection and the thickness of the weld. The staff determined that the acceptance criteria for the examination of the welds is sufficient to ensure that the package components and welds provide adequate gamma shielding.

Based on review of the statements and representations in the application, the NRC finds that the acceptance tests and maintenance program have been adequately described and meet the requirements of 10 CFR Part 71.

8.4 Alternate Lead Pour Procedure The applicant provided an alternate lead pour procedure in SAR Section 8.4.3, Description of Lead Pour Procedures (Alternate Method). In SAR Section 8.4.3.2, Lead Pour Operations, the applicant clarified in that the allowable heat up rate is the same as the standard lead pour procedure. The applicant also described in Section 8.4.3.3, Cooldown Following Lead Pour, of the SAR a minimum inlet water temperature and a maximum rate of rise for the water level.

These cooldown values are both necessary to limit the development of any radial or axial thermal gradients greater than what was analyzed as part of the response to RAI 8.6, (ADAMS Accession No. ML18145A236), in addition, NAC provided a sensitivity study in Calculation Package No. 30045-3000 Revision 0, Thermal Simulation of NAC-STC Cask Cooldown.

Based on the staffs review of the description of the sensitivity study and because the parameters were included in Sections 8.4.3.2 and 8.4.3.3 of the SAR, the staff finds the alternate lead pour procedure acceptable. The staff reviewed the description of the alternate lead pour procedures. The staff confirmed that the applicant adequately described the alternate method for lead pouring including the lead specifications, allowable lead temperatures, procedures to avoid the formation of voids, heating and allowable temperatures for the NAC-STC body weldment, cooldown procedures following the lead pour and lead pour documentation. The staff determined that the alternate lead pour description is adequate to ensure that the lead pouring operation will result in lead shielding with satisfactory performance.

8.5 Change to Periodic Thermal Test The amendment requested to a change to the periodic thermal test described in Sections 3.4.8, Assessment criteria for the package passive heat rejection system, and 8.2.6, Periodic thermal test, of the SAR. The thermal test that is performed on each fabricated packaging prior to acceptance at the factory remains unchanged as described in Section 8.1.6, Thermal test, of the SAR. Sections 8.2.6, Post-fabrication thermal test, and 8.2.7, Miscellaneous, of the SAR were modified to describe a visual inspection of the surface of the package that includes a visual inspection on the radial neutron shield to confirm there is no change of the heat transfer capability of the packaging.

The applicant stated in the response to RAI 8.1, (ADAMS Accession No. ML18289A389) that thermal fins are welded to the cask body outer shell using a full penetration groove weld, and the thermal fins are also welded to the neutron shield shell using full penetration groove welds or a single full penetration double-bevel weld that are all examined using dye penetrant testing (PT). The applicant also stated in the response to RAI 8.1 that the thermal cycling would not impose significant stresses, and normal operations would not cause these connections to fail.

The applicant concluded that any impact capable of imposing loads resulting in deformation, or weld failure would be visible by the general visual inspection described in Section 8.2.7 of the SAR. Section 8.2.7 of the SAR states, The radial neutron shield shall be visually inspected prior to each fuel loading. Any crack, gouge, or gross deformation that could indicate damage of the heat transfer fins shall be cause for rejection of the cask for use until approved maintenance and/or repair activities have been acceptably completed.

Section 8.2.6 of the SAR also describes that if during handling or transport operations, the packaging experiences an adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield, a thermal test will be performed on the operational packaging in accordance with Section 8.1.6 of the SAR. Based on the visual inspection described in Sections 8.2.6 and 8.2.7 of the SAR, and that any obvious damage to the neutron shield would result in a thermal test being performed in accordance with Section 8.1.6 of the SAR, the staff finds the visual inspection acceptable.

8.6 Findings

The staff reviewed and evaluated the revised acceptance tests. Based on the statements and representations in the application, the staff concluded that the revised acceptance tests meet the requirements of 10 CFR Part 71, and that they are adequate to assure the package will be constructed in a manner consistent with its evaluation for approval. Further, the certificate is conditioned to specify that the package must be prepared for shipment and operated in accordance with the Acceptance Tests and Maintenance Procedures in Chapter 8 of the SAR, as amended.

9.0 CONDITIONS The staff made editorial changes to improve the readability of the CoC. The CoC includes the following condition(s) of approval:

The following new and revised drawings were incorporated into the certificate of compliance:

The revised drawings showing the transport packaging include:

423-800, Sheets 1-3, Rev. 20P and 20NP Cask Assembly - NAC-STC Cask 423-802, Sheets 1-7, Rev. 26 Cask Body - NAC-STC Cask 423-803, Sheets 1-2, Rev. 15 423-804, Sheets 1-2, Rev. 12 Details - Inner Lid, NAC-STC Cask 423-805, Sheets 1-2, Rev. 9 Details - Inner Lid, NAC-STC Cask 423-806, Sheets 1-2, Rev. 14 Port Coverplate Assy - Inner Lid, NAC-STC Cask 423-807, Sheets 1-3, Rev. 6 Assembly, Port Cover, NAC-STC Cask 423-811, Sheets 1-2, Rev. 13 Details -NAC-STC Cask 423-812, Rev. 7 Nameplates - NAC-STC Cask 423-859, Rev. 1 Attachment Hardware, Balsa Limiters, NAC-STC 423-870, Rev. 8 Fuel Basket Assembly, PWR, 26 Element, NAC-STC Cask 423-874, Rev. 3 Heat Transfer Disk, Fuel Basket, PWR, 26 Element, NAC-STC Cask 423-878, Sheets 1-2, Rev. 5 Alternate Tube Assembly, NAC-STC Cask 423-880, Rev. 3P Shielded Thermal Shunt Assembly, NAC-STC Cask 423-900, Rev. 9 Package Assembly Transportation, NAC-STC Cask 423-209, Rev. 2 Impact Limiter Assy - Upper, NAC-STC Cask 423-210, Rev. 2 Impact Limiter Assy-Lower, NAC-STC Cask 423-257, Rev. 3 Balsa Impact Limiter, Upper, NAC-STC Cask and 423-258, Rev. 3 Balsa Impact Limiter, Lower, NAC-STC Cask.

New drawing:

423-927, Rev. 1P & 2NP Shield Ring Assembly, NAC-STC Cask The description in 5.(a)(2) was revised to clarify that SA-705 Type 630, H1150 is 17-4PH stainless steel and added a description of the shield ring.

Content 5.(b)(1)(i)(3) was added for undamaged 17x17 advanced fuel assembly PWR LBU fuel assemblies that meet the fuel assembly criteria for Framatome-Cogema 17x17 fuel listed in Table 1 for content 5.(b)(1)(i)(1). Content 5.(b)(1)(i)(3) also includes reference to new cooling table data for LBU fuel and adds reference to the shield ring assembly. Added loading Table 6 in the certificate.

Content 5.(b)(1)(i)(4) added for undamaged 17x17 advanced fuel assembly PWR HBU (i.e., assembly average burnup exceeding 45 GWd/MTU) fuel assemblies that meet the fuel assembly criteria for Framatome-Cogema 17x17 fuel listed in Table 1 for content 5.(b)(1)(i)(1).

Added loading Tables 7, 8, and 9 in the certificate.

Contents 5.(b)(1)(i)(2) was edited for clarity, and consistency with 5.(b)(1)(i)(4).

Added 5.(b)(2)(i)(3) to state the maximum quantity of material per package for LBU fuel.

Added 5.(b)(2)(i)(4) to state the maximum quantity of material per package for HBU fuel and LBU comingled loads.

Table numbers 6-8 were renumbered to 10-12.

The references section was updated to include the applications for the two amendments dated March 16, 2017 and December 8, 2017, as supplemented on July 17, 2017, September 20, 2017, March 6, 2018, June 5, 2018, June 21, 2018, July 18, 2018, August 21, 2018, September 19, 2018, October 12, 2018, and October 26, 2018.

10.0 CONCLUSION

S Based on the statements and representations contained in the application, as supplemented, and the conditions listed above, the staff concludes that the design has been adequately described and evaluated, and the Model No. NAC-STC package meets the requirements of 10 CFR Part 71.

Issued with Certificate of Compliance No. NAC-STC, Revision No. 19.

11/7/18

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