ML21057A160

Transmittal Letter and SER: Revalidation Recommendation for the Japanese Certificate of Approval No. J/2009/AF-96, Revision 1, Model No. GP-01

ML21057A160
Person / Time
Site:	07103098
Issue date:	03/01/2021
From:	John Mckirgan Storage and Transportation Licensing Branch
To:	Boyle R US Dept of Transportation, Radioactive Materials Branch
NGSantos - NMSS/DFM/STL 301.415.6999
Shared Package
ML21057A159	List: ML21057A160
References
EPID L-2020-LLA-0107, EPID L-2021-DOT-0000
Download: ML21057A160 (39)

Text

SorryorryUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 1, 2021.

Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 400 Seventh Street, S.W.

Washington, D.C. 20590

SUBJECT:

REVALIDATION RECOMMENDATION FOR THE JAPANESE CERTIFICATE OF APPROVAL NO. J/2009/AF-96, REVISION 1, MODEL NO. GP-01 PACKAGE (DOCKET NO. 71-3098)

Dear Mr. Boyle:

By letter dated May 11, 2020 [Agencywide Documents Access and Management System (ADAMS) Accession No. ML20143A100], and as supplemented on August 5, 2020 (ADAMS Package Accession No. ML20231A505) and January 12, 2021 (ADAMS Package Accession No. ML21013A481), the U.S. Department of Transportation requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the Japanese Approval Certificate Number J/2009/AF-96, Revision 1, Model No. GP-01 transport package, and make a recommendation concerning the revalidation of the package for import and export use against the requirements found in International Atomic Energy Agency Specific Safety Requirements No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, 2012 Edition.

Based upon our review of the statements and representations contained in the application and its supplements, and for the reasons stated in the enclosed safety evaluation report, we recommend revalidation of the Japanese Certificate of Approval No. J/2009/AF-96, Revision 1, for the Model No. GP-01 package, with the following additional condition:

Transport by air of the Model No. GP-01 is not authorized.

If you have any questions regarding this matter, please contact me or Norma García Santos of my staff at (301) 415-6999.

Sincerely, John McKirgan, Branch Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3098 EPIDs L-2020-LLA-0107 and L-2021-DOT-0000

Enclosures:

1. Safety Evaluation Report

2. Enclosure 1 to E-56661 Japanese Certificate of Approval No. J/2009/AF-96, Revision 1 in English (ADAMS Accession No. ML2014A100)

John B.

McKirgan Digitally signed by John B.

McKirgan Date: 2021.03.01 14:09:13

-05'00'

(Transmittal letter and SER): ML ADAMS Accession No. (Japanese Certificate of Competent Authority): ML

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3098 Model No. GP-01 Package Certificate of Approval No. J/2009/AF-96 Revision 1

Table of Contents Page

SUMMARY

................................................................................................................................ 1 1.0 GENERAL INFORMATION............................................................................................ 1 1.1 Package Description........................................................................................... 1 1.1.1 Packaging............................................................................................... 1 1.1.2 Contents.................................................................................................. 3 1.2 Drawings............................................................................................................. 3 2.0 STRUCTURAL EVALUATION........................................................................................ 4 2.1 Description of Structural Design.......................................................................... 4 2.2 Mechanical Analysis........................................................................................... 4 2.2.1 Stacking Test.......................................................................................... 4 2.2.2 Lifting Test............................................................................................... 5 2.2.3 Penetration Test...................................................................................... 5 2.2.4 Tie-Down Test......................................................................................... 6 2.3 Structural Evaluation under Normal and Accident Conditions of Transport.......... 6 2.3.1 Drop Tests............................................................................................... 6 2.3.2 Water Immersion Test............................................................................. 6 2.4 Evaluation Findings............................................................................................. 6 3.0 MATERIALS EVALUATION............................................................................................ 7 3.1 Design Criteria.................................................................................................... 7 3.1.1 Codes and Standards.............................................................................. 7 3.1.2 Weld Design and Inspection.................................................................... 7 3.2 Material Properties.............................................................................................. 7 3.2.1 Mechanical Properties............................................................................. 7 3.2.2 Thermal Properties.................................................................................. 8 3.2.3 Fracture Toughness................................................................................ 8 3.2.4 Criticality Control..................................................................................... 8 3.3 Corrosion, Chemical Reaction, and Radiation Effects......................................... 9 3.3.1 Corrosion Resistance/Content Reactions................................................ 9 3.3.2 Protective Coatings................................................................................. 9 3.3.3 Radiation Effects..................................................................................... 9 3.4 Content Integrity................................................................................................. 9 3.4.1 Fresh Fuel/Contents of Packaging........................................................... 9 3.5 Component-Specific Reviews............................................................................10 3.5.1 Gasket....................................................................................................10 3.5.2 Rod Bolts................................................................................................10 3.5.3 Shock Absorbers....................................................................................11 3.6 Evaluation Findings............................................................................................11 4.0 THERMAL EVALUATION..............................................................................................11 4.1 Description of the Thermal Design.....................................................................12 4.2 Material Properties and Component Specifications............................................13 4.2.1 Material Properties....................................................................................13 4.2.2 Component Specifications.........................................................................13 4.3 General Considerations.....................................................................................13 4.3.1 Evaluation by Test.....................................................................................13 4.3.2 Evaluation by Analysis............................................................................14 4.3.3 Contents Decay Heat.............................................................................15 4.3.4 Summary Tables of Temperatures.........................................................15

ii 4.3.5 Margins of Safety...................................................................................16 4.4 Thermal Evaluation under Normal Conditions of Transport................................16 4.4.1 Heat and Cold........................................................................................16 4.4.2 Maximum Normal Operating Pressure....................................................16 4.4.3 Maximum Thermal Stresses...................................................................16 4.5 Thermal Evaluation under Accident Conditions of Transport (ACT)...................16 4.5.1 Initial Conditions.....................................................................................16 4.5.2 Fire Test Conditions...............................................................................17 4.5.3 Maximum Temperatures and Pressure...................................................17 4.5.4 Maximum Thermal Stresses...................................................................17 4.5.5 Analyses Details.....................................................................................17 4.6 Evaluation Findings............................................................................................17 5.0 CONTAINMENT EVALUATION.....................................................................................17 5.1 Description of the Containment System.............................................................18 5.1.1 Temperatures (Highest and Lowest).......................................................18 5.1.2 Leaktight Analysis..................................................................................19 5.1.3 Normal Conditions of Transport..............................................................19 5.1.4 Accident Conditions of Transport (ACT).................................................20 5.1.5 Inspections and Maintenance.................................................................21 5.2 Evaluation Findings............................................................................................21

6.0 CRITICALITY EVALUATION

.........................................................................................21 6.1 Description of Criticality Design..........................................................................22 6.2 Package Contents..............................................................................................23 6.3 Considerations for Criticality Evaluations...........................................................23 6.3.1 General Model Configuration Considerations.........................................23 6.3.2 Undamaged Package Array Model.........................................................25 6.3.3 Damaged Package Array Model.............................................................25 6.3.4 Material Properties.................................................................................28 6.3.5 Analysis Methods and Nuclear Data section...........................................28 6.3.6 Demonstration of Maximum Reactivity and Criticality Safety Index.........29 6.3.7 Confirmatory Calculations......................................................................29 6.4 Benchmark Evaluation.......................................................................................29 6.5 Operations, Acceptance Tests and Maintenance Programs Related to Criticality Safety................................................................................................................31 6.6 Evaluation Findings............................................................................................31 7.0 QUALITY ASSURANCE................................................................................................32 7.1 Staffs Evaluation of the Quality Assurance Program.........................................32 7.2 Evaluation Findings............................................................................................32

8.0 REFERENCES

..............................................................................................................32 CONDITIONS...........................................................................................................................33 CONCLUSION..........................................................................................................................33

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3098 Model No. GP-01 Package Certificate of Approval No. J/2009/AF-96 Revision 1

SUMMARY

By letter dated May 11, 2020 (DOT, 2020a), and as supplemented on August 5, 2020 (DOT, 2020b) and January 12, 2021 (DOT, 2021), the U.S. Department of Transportation requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the Japanese Approval Certificate Number J/2009/AF-96, Revision 1, Model No. GP-01 transport package, (DOT, 2020a) and make a recommendation concerning the revalidation of the package for import and export use.

The NRC reviewed the information provided to the DOT by TN Americas LLC (Orano or the applicant) in its application for the Model No. GP-01 package and its supplements against the regulatory requirements of the International Atomic Energy Agency (IAEA) Safety Standard Series No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, 2012 Edition (IAEA, 2012). Based on the statements and representations in the information provided by DOT and the applicant, the staff recommends the revalidation of the Japanese Certificate of Approval No. J/2009/AF-96, Revision 1, Model No. GP-01 package, (DOT, 2020a) for the contents included in Section 1.1.2, Contents, of this safety evaluation report (SER), with the added condition described at the end of this SER.

1.0 GENERAL INFORMATION The application includes a description of the Model No. GP-01 package (DOT, 2020a and DOT, 2020b). This section of this SER also includes a brief description of the package.

1.1 Package Description The GP-01 package is a Type A fissile package designed to transport uranium dioxide fuel pellets. TN Americas LLC, on behalf of TN International (TNI), submitted a description of the design of the GP-01 package as part of the Certification for Approval of Package Design for Transport of Radioactive Materials (Certificate) for the GP-01 package, Identification Mark:

J/2009/AF-96 (Rev. 1), dated August 3, 2018 (NFK-MPCF-1808001), issued by the Competent Authority of Japan, the Nuclear Regulation Authority (NRA) (DOT, 2020a).

1.1.1 Packaging Chapter I and Figures I-1 through I-11 of the SAR include a description of the Model No. GP-01 package general arrangement. The packaging consists of an outer container, referred to as the outer receptacle; an inner container, referred to as the inner receptacle; and pellet box

2 assemblies, of which there are two types that hold different amounts of fissile contents.

1.1.1.1 Outer receptacle The outer receptacle has a multi-caisson-shaped double structure composed of the following components (described below in further detail):

(1) stainless steel frames, (2) inner/outer stainless steel plates (where the voids are filled with blocks of ceramic fiber insulating material to ensure heat resistance), and (3) spacers fabricated from rubber.

Also, rubber spacers and an O-ring provide a seal to the receptacle on the flange surface.

The outer container body and lid are of a double wall construction. The walls are made of stainless steel. Stainless steel frames are between the walls and support the containers shape and separate the cavity between the inner and outer walls into multiple cavities within which there is ceramic insulation material. Aluminum honeycomb items are also attached to the interior surfaces of the outer container, its walls, base and lid. These items provide positioning and impact absorption for the inner container. In addition, the outer container has various rubber items. These items include silicone rubber at the interface of the body flange and lid and anti-vibration pads attached to the aluminum honeycomb on the base of the interior for the inner container to sit on. The package also has legs which extend into recesses of the package lid, when packages are stacked on each other. Each of the outer faces of the outer receptacle has a stainless steel fusible plug installed with solder [Japanese Industrial Standards (JIS) Z 3282]

used as the fusible material. Chromium-molybdenum rod bolts with stainless steel nuts keep the outer receptacle stainless steel lid (same structure as body) joined to the body of the outer receptacle.

1.1.1.2 Inner receptacle The inner container body and lid are of single wall (i.e., single-caisson) construction. They are made of stainless steel plates. Six borated stainless steel plates [with 1 weight percent (wt. %)

boron] are attached to the inside of the walls of the container using an adhesive, and two borated stainless steel plates (also with 1 wt. % boron) are fixed by bolted attachments in the center of the container cavity as a partition that divides the cavity into two equal spaces, into each of which a pellet box assembly can be placed. There are some rubber components in the inner container as well that serve to position the pellet box assemblies within the container cavity and prevent damage to the borated steel plates from insertion and removal of the pellet box assemblies. The lid of the inner receptacle is joined to the body of the inner receptacle by means of chromium-molybdenum rod bolts with stainless steel nuts.

1.1.1.3 Pellet Box Assemblies The pellet box assemblies consist of four stainless steel walls, a bottom and a top stainless steel plate and intermediate plates, or partitions, of borated stainless steel (with 1 wt. % boron).

These plates are alternately stacked with pellet boxes and are at fixed separation by attachment to posts that support the stacks of plates, partitions and pellet boxes. The assemblies also have rubber and steel components to maintain the pellet boxes position vertically and horizontally

3 between the plates and partitions. There are two types of pellet box assemblies. The assemblies differ in the stack configuration, and pellet box sizes they accommodate. Only one type of assembly is loaded into a specific package.

1.1.2 Contents Table 2 of the Japanese Certificate of Competent Authority J/2009/AF-96, Revision 1, (ADAMS Package Accession No. ML20141L696) includes a brief description of the allowable contents of the package. The content of the package consists of solid, unirradiated pellets of uranium oxides (UO2, UO3 and U3O8) or uranium oxides mixed with gadolinia. Specifically, as described in the Japanese certificate J/2009/AF-96 (Rev. 1), and further defined in Chapter I-D. of the SAR (NFK-MPC-1801024), the approved contents for this package include:

two assemblies of pellet storage boxes which contain pellets of uranium oxides or pellets of uranium oxides with gadolinia.

The application and Japanese certificate indicate that the term uranium oxide includes UO2, UO3, and U3O8. However, the applicant intends that UO3 and U3O8 be only impurities in UO2 pellets and not distinct pellet materials (i.e., the package will not be used to ship UO3 pellets and U3O8 pellets) with or without gadolinia mixed in the pellets.

The maximum enrichment of uranium is 5 wt. % of uranium-235 (235U) for use in light water reactors. Section 3.4.1 of this SER include additional descriptions of the packages contents.

The pellets are contained in stacks of trays in pellet boxes that are themselves stacked in their respective pellet box assemblies. There are two types of pellets storage box assemblies. They are the following:

a)

Assembly A consisting of twelve (12) pellet storage boxes, which can store up to 11 [kilograms] kg of UO2 per pellet box for a maximum assembly capacity of 132 kg of UO2; and b)

Assembly B consisting of five (5) pellet storage boxes, which can store up to 20 kg of UO2 per pellet box for a maximum assembly capacity of 100 kg of UO2.

The package can only ship one type of pellet storage box assembly. In other words, a package may not contain an assembly A and an assembly B together in that package.

1.2 Drawings The staff reviewed the SAR and drawings in document No. NFK-MPCF-1803001-2, Drawings of GP-01, (ADAMS Package Accession No. ML20141L696) and verified that the applicant provided an adequate description of the following:

1) component safety functions,

2) materials of construction,

3) dimensions and tolerances, and

4) fabrication (welding) specifications.

4 In Table I-1, Major Materials for the Packaging Components, of the SAR, the applicant specifies that the austenitic stainless steels used in the structural components of the Model No.

GP-01 transportation package conform, typically, to Japanese standards JIS G 4304 or JIS G 4305, and that several potential stainless steel types could be used within those standards. In addition, Table II-A-3, Mechanical Properties of Major Structural Materials, of the SAR includes the mechanical property requirements for specific austenitic stainless and chrome-molybdenum steels. Therefore, the staff finds that the applicant provided acceptable information in the SAR and associated drawings to describe the packaging materials.

2.0 STRUCTURAL EVALUATION The purpose of the structural evaluation is to verify that the structural performance of the package meets the requirements of IAEA SSR-6, 2012 Edition. A summary of the staffs structural evaluation is provided below.

2.1 Description of Structural Design The Model No. GP-01 is designed to protect the radioactive materials during normal conditions of transport (NCT) and in accident conditions of transport (ACT) as required by IAEA SSR-6, 2012 Edition.

Section 1.1 of this SER includes a general description of the package. Structural design features of the package include:

1) outer receptacle consisting of stainless-steel structural elements including frame elements, inner and outer plates with insulation,

2) aluminum honeycomb elements to serve as shock absorber between the outer and inner receptacle, and

3) inner receptacle serving as the confinement system, consisting of stainless-steel plates provided with rubber spacer and an O-ring.

The applicant provided the general assembly figures of the GP-01 transportation package in Chapter II, Safety Analysis of Nuclear Fuel Package, of the SAR. The staff reviewed the figures and diagrams for completeness and accuracy and finds that the applicant adequately described the relevant details of the major components of the GP--01 package. Section 1.0 of this SER includes staffs additional evaluation of the package drawings.

2.2 Mechanical Analysis The following sections include a discussion of the information provided by the applicant related to the mechanical analysis of the GP-01.

2.2.1 Stacking Test The IAEA SSR-6, 2012 Edition, requires subjecting the GP-01 to the approach that may result in the maximum compressive stress on the package:

5 a) 5 times the weight of the package, or b) 13 kilopascals (kPa) [1.9 pounds force per square inch (lbf/in2)] times the vertical projected area of the package.

The applicant selected the first approach as the most challenging to the structural integrity of the package. The applicant performed calculations to demonstrate the acceptability of the package using the selected approach.

In Section A.5.4 of the SAR, the applicant calculated the applicable stacking pressure loads for the outer receptacle of the package based on the IAEA SSR-6, 2012 Edition. The load value of 5 times the mass of the package was distributed uniformly to the eight channel elements that provide the support for the corners of the outer receptacle. The applicant compared the resulting pressure to the buckling capacity of the channel skeleton elements in the four corners of the outer receptacle. The applicant concluded that the capacity of the channel elements was significantly higher than the superimposed load, since a margin of safety of 12.7 was calculated between the buckling capacity and the load pressure. The applicant concluded that the GP-01 package meets the requirements of the stacking test in IAEA SSR-6, 2012 Edition.

The staff reviewed the analysis and test results submitted by the applicant. Based on the information provided by the applicant, the staff finds that the package meets the requirements prescribed in IAEA SSR-6, 2012 Edition.

2.2.2 Lifting Test In Section A.4.4 of the SAR, the applicant described and provided diagrams of the lifting devices for the outer and inner receptacle. The material of the lifting attachments is stainless steel and are welded to frames of the outer and inner receptacle. The applicant designed all lifting attachments to support three times the weight of the package, which is accomplished by using four steel wires. The applicant evaluated the stresses that will develop in the lifting attachments from lifting operations of both the receptacles. The margin of safety between the calculated stresses and the allowable stresses of the lifting attachments was greater than 1.4 for all lifting scenarios.

The staff reviewed the analysis and test results submitted by the applicant. Based on the information provided by the applicant, the staff finds that the applicant evaluated all lifting attachments to ensure that the package meets the requirements prescribed in IAEA SSR-6, 2012 Edition.

2.2.3 Penetration Test In Appendix 1 to Chapter II-A of the SAR, the applicant provided the results of the penetration tests required by IAEA SSR-6, 2012 Edition. The applicant selected specimen orientations and drop locations to ensure that the most severe drop conditions were covered. The tests were carried out as part of the drop tests and were located on the central area of the lateral sides of the package which are not directly supported by the frames, the tightening rod bolts, and the fusible plug. No major damages were reported on any of the three test locations and only dents were noted.

6 The staff reviewed the test results submitted by the applicant for the penetration tests and finds that, in aggregate, all of the tests that were completed meet the requirements of IAEA SSR-6, 2012 Edition.

2.2.4 Tie-Down Test The package does not incorporate any design features that are used as a tie-down device. The applicant stated that the package to be transported should be tied down only with steel wires or dedicated tie-down attachments on a vehicle or a transport container. Therefore, the requirement of IAEA SSR-6, 2012 Edition, to evaluate tie-down attachments is not applicable to this package.

2.3 Structural Evaluation under Normal and Accident Conditions of Transport The following sections include a summary of the information provided by the applicant related to the structural analysis under NCT and ACT for the Model No. GP-01.

2.3.1 Drop Tests The applicant performed a series of drop tests with various impact configurations as described in Sections A.5.3 and A.9.2 as well as Appendix 1 to Chapter II-A of the SAR. The applicant used two prototype packages with the same structural features of an actual GP-01 package for the tests. Prototype No. 1 was used to determine the critical orientation to drop Prototype No. 2 under NCT and ACT. The tests included free drop tests from 1.2 meters (m), followed by penetration tests under NCT, a 9-m drop test, and a 1-m drop test onto a target. The drop test orientations were chosen to cause maximum damage to the package, where the downward corner drop was considered the most critical. With the results of the testing campaign, the applicant concluded that the outer receptacle deforms and that the inner receptacle moves inside the outer receptacle. Consideration of the movement of the inner receptacle based on the deformations seen from the test results are described in Table II-A-12 of the SAR and were used as considerations for the criticality analysis. For the inner receptacle, the applicant stated that, based on the test results, the nuclear fuel material remains in the pellet storage boxes and inside the inner receptacle.

The staff reviewed the analysis and test results submitted by the applicant and finds that these tests and the analysis, in aggregate, meet the requirements IAEA SSR-6, 2012 Edition, for drop tests under NCT and ACT.

2.3.2 Water Immersion Test In Section A.9.2.4 of the SAR, the applicant noted that the criticality analysis considers water infiltration into the inner receptacle. Therefore, the packaging structure is not subject to the loading of the water immersion test.

2.4 Evaluation Findings

Based on the review of the structural analysis and test results for the GP-01, the staff finds that the GP-01 meets the requirements of IAEA SSR-6, 2012 Edition.

7 3.0 MATERIALS EVALUATION The staff reviewed the application with respect to compliance with the IAEA SSR-6 regulations, 2012 Edition, consistent with the approval under the Japanese Competent Authority Certificate of Approval J/2009/AF-96, Revision 1 (DOT, 2020a). The staff conducted a review of the package to determine, with reasonable assurance, that the performance of the materials used to build the package components is adequate per the regulations in IAEA SSR-6, 2012 Edition.

The staff also reviewed the application to ensure that standards, specifications, and acceptance tests for fabrication of the package are properly defined and reasonable for supporting the intended functions of the package components under the loads and environments required for evaluation per IAEA SSR-6, 2012 Edition.

3.1 Design Criteria 3.1.1 Codes and Standards Chapter II-A.1.2 of the SAR includes a description of applicable design standards. The analysis was carried out to evaluate the structure of the Type A fissile package under NCT and ACT.

The applicant referenced the Japan Society of Mechanical Engineers (JSME) S NC1-2001 design and construction standard for Nuclear Facilities for Electric Power Generation. The staff notes that the American Society of Mechanical Engineers (ASME) Boiler Pressure and Vessel (BP&V) Code Section III, Division 1, is the counterpart to JSME S NC1. Table II-A-1 of the SAR includes a description of the design criteria for the structural analysis of the package including reference drawings, materials, design temperatures, and design loads. The Model No. GP-01 structures, systems, and components (SSCs) are typically fabricated with austenitic stainless steel (type SUS 304) and chromium-molybdenum steel (SCM 435 H) to JIS G 4304/JIS G 4305 and JIS G 4052, respectively. In addition, the borated stainless steel is fabricated following the American Society for Testing and Materials (ASTM) A887-89 standard. The staff finds that the applicant adequately referenced package materials codes and standards, which provide materials chemistry, mechanical property, and fabrication requirements. Therefore, the staff finds the packages material codes and standards acceptable.

3.1.2 Weld Design and Inspection Chapter I-C(17) of the SAR includes a description of the Model No. GP-01 package welding.

The applicant referenced the Japan Welding Society, Manual for Welding, 3rd revised edition.

In addition, per a request for additional information, the applicant provided JIS Z 3821, Standard qualification test and acceptance requirements for welding technique of stainless steel, which specifies the standard qualification test and acceptance requirements for various welding techniques and combined welding of stainless steels. The staff reviewed the weld design and finds that the applicant adequately identified weld and inspection standards.

Therefore, the staff finds the Model No. GP-01 package weld design and inspection to be acceptable.

3.2 Material Properties 3.2.1 Mechanical Properties Chapter II-A.3 of the SAR includes a description of mechanical properties of packaging materials. The applicant stated that metallic materials for the packaging include austenitic stainless steel (equivalent to SUS 304). Table II-A-3 of the SAR includes a description of the

8 mechanical properties of the major materials, which comprise the Model No. GP-01 packaging.

The staff reviewed the temperature-dependent mechanical properties of stainless steels used in the applicants mechanical calculations and confirmed that the properties are consistent with those in the technical literature (e.g., ASME Code Section II, data sheets, handbooks, etc.).

Therefore, the staff finds the applicant adequately identified the temperature dependent properties of the stainless steels used in the Model No. GP-01 package.

3.2.2 Thermal Properties Tables II-B-1 through II-B-4 of the SAR identified the thermophysical properties of the materials to be contained in the Model No. GP-01 package inner receptacle (containment), which were used for the thermal analysis of the package. The staff evaluated the applicants thermal properties of the materials credited in the thermal analysis and determined that the thermal properties (e.g., thermal conductivity, specific heat, density, etc.) are consistent with those in the technical literature. Therefore, the staff finds that the applicant adequately identified the thermal properties of the materials used in the Model No. GP-01 package. The staff notes that the mechanical and thermal properties of the chromium-molybdenum and aluminum are further evaluated in Section 3.5 of this SER.

3.2.3 Fracture Toughness Chapter II-A.4.2 and Appendix 2 of the SAR include a description of the low-temperature properties of the Model No. GP-01 package materials. The staff notes that metallic materials for the packaging include austenitic stainless steel and aluminum alloy. The applicant stated that these materials do not lose their strength or toughness in an environment kept at -40°C.

Figures II-A.App2-1 to II-A.App2-2 of the SAR include the low-temperature tensile and impact characteristics for austenitic stainless steel. In addition, Figure II-A.App2-5 of the SAR includes tensile characteristics at low temperatures for aluminum. The staff finds that the applicant has adequately considered fracture toughness behavior of the above stated materials used in the Model No. GP-01 package design. Therefore, the staff finds the austenitic stainless steel and aluminum alloy used in the package construction to be acceptable. The staff notes that fracture toughness of the chromium-molybdenum is evaluated in Section 3.5 of the SER.

3.2.4 Criticality Control Chapter I-C-V and Figure I-6 of the SAR include a description of the neutron absorbers and locations. The applicant stated that the neutron absorbers are borated stainless steel plates, six of which are affixed to the inner receptacles walls inner surfaces with an inorganic adhesive and two of which are stacked together and fixed in position by attachment bolts in the center of the inner receptacle cavity, dividing the cavity into two halves. In addition, all the borated stainless steel plates are fabricated to ASTM A887-89, Types 304B, and 304B4, with boron concentration within the range 1.0 to 1.24 wt. % and a density of 7.8 grams per cubic centimeters (g/cm3). Spacers of neoprene rubber are applied to the surfaces of all the neutron absorbers to prevent friction wear during operation. The applicant stated that if the temperature of the borated stainless steel plates is conservatively assumed to reach 170°C in the inner receptacle, no deformation or deterioration will occur that should be considered in the criticality evaluation. The staff verified that the applicant appropriately described the density and geometry of the Model No. GP-01 neutron absorber materials. Therefore, the staff finds the material properties description of the criticality control components to be acceptable. Section 6.0 of the SER includes the staffs criticality evaluation for the Model No. GP-01 including the uses of criticality control material properties in the criticality analysis.

9 3.3 Corrosion, Chemical Reaction, and Radiation Effects 3.3.1 Corrosion Resistance/Content Reactions Chapter II-A.4.1 of the SAR includes a description of the chemical and galvanic reactions of the Model No. GP-01 packaging materials. Table II-A-4 of the SAR includes a list of the packaging materials that stay in contact with each other or with parts of the package contents. The applicant stated that none of the contact combinations for these materials produce hazardous chemical or galvanic reactions. The stainless steel, borated stainless steel, neoprene rubber, and organic polymeric materials (polyethylene, polyvinyl chloride and urethane) have chemically stable properties, do not react with each other, nor will these materials give rise to corrosion.

The staff notes that galvanic corrosion between the aluminum honeycomb and the welded stainless steel plates is not expected because water is effectively sealed off under NCT. The applicant stated that the Model No. GP-01 package is not required to be subjected to the enhanced water immersion tests. However, the package is subject to water spray tests, which are described in Chapter II-A.5.2 of the SAR. No entry of sprayed water occurs in the package.

The applicant stated that the threaded portion of the fusible plug has a high leak-tightness ensured by an O-ring. The applicant stated that the external surfaces of the packaging materials do not deteriorate with sprayed water, and water will not enter the heat insulator zones. In addition, the heat insulator is made of ceramic fiber and is free from deterioration by water. The outer receptacle body flange is designed such that the inner side is higher than the outer side, therefore, water (e.g., rainwater) is prevented from entering the outer receptacle through the interface between the body and the lid. The staff notes that periodic visual inspections of the package can identify corrosion should it arise.

The staff finds that the applicant appropriately accounted for chemical and galvanic reactions of the Model No. GP-01 unirradiated-fuel package and that no credible corrosion, galvanic, or other adverse reactions will exist during NCT. Therefore, the staff finds acceptable the resistance of the packages materials to corrosion and content-related chemical and galvanic reactions.

3.3.2 Protective Coatings None 3.3.3 Radiation Effects None 3.4 Content Integrity 3.4.1 Fresh Fuel/Contents of Packaging Chapter I-D and Figure I-8 of the SAR include a description of the packaging contents as two assemblies of pellet storage boxes, which contain cylindrical ceramic solid pellets formed from unirradiated uranium oxides or uranium oxides mixed with gadolinia by press-molding and sintering. The pellets (density of 8 to 11 g/cm3) are lined up on stainless steel corrugated plates in the stainless steel pellet storage box. In addition, spacers of an organic polymeric material (neoprene rubber or urethane foam) are inserted between corrugated plates, as required, to minimize the adverse effect of vibration during NCT. The applicant stated that pellet storage

10 boxes are sometimes sealed in a plastic (polyethylene) bag for operational reasons. In such cases, plastic bags are welded by means of a sealer or are protected with pieces of polyvinyl chloride adhesive tape. The stack of pellet storage boxes is fixed with stainless steel nuts at the threaded top of pillars. In addition, the pellet storage box assembly is made of stainless steel except for the partition plates between pellet storage boxes in an assembly, which are borated stainless steel plates, and a few rubber parts.

Chapter I-D(3) of the SAR describes the ceramic pellets as chemically stable because these a) do not react with other packaging materials, b) do not provoke electro-chemical reactions, and c) will not present risks of corrosion.

The staff finds that the applicant appropriately described the fuel contents, density, and geometry. The staff also finds that the fuel contents are not subject to adverse chemical reactions and will not reconfigure under NCT and ACT of the Model No. GP-01 package.

Therefore, the staff finds the package contents to be acceptable.

3.5 Component-Specific Reviews 3.5.1 Gasket In Chapter I-C(6) and (15) of the SAR, the applicant noted that an O-ring provides sealing for the inner receptacle (containment boundary) on the flange surface to prevent water ingress to the fuel pellet contents. The packages configured with the Model No. GP-01 packaging are classified as Type A fissile transport packages and do not need to meet the regulatory requirements for leak-tightness under ACT. However, the entire finished inner receptacle is inspected for leak-tightness in water at least 1 m in depth or under an equivalent hydraulic pressure for at least one hour. The silicone rubber O-ring serves as a gasket installed (engaged) in a groove [13 millimeters (mm) wide (W) by 7mm diameter (D)]. The O-ring has a service temperature range of -50 to 180°C and no serious deterioration (cracking or fracture) occurred when tested at 225°C for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The gaskets tensile strength is 3.4 MPa, hardness is 48 to 60 (measured with a Type A durometer), and elongation is 200%. Table II-B, App2-6, of the SAR includes a description of the rubbers thermophysical properties used for thermal analyses. The staff verified that the mechanical and thermal properties of the gasket are consistent with those from open literature (e.g., data sheets, handbooks, etc.). The staff finds the applicant appropriately described the gasket material used in the construction of the Model No. GP-01 package. Therefore, the staff finds the gasket material to be acceptable.

3.5.2 Rod Bolts Chapter I-C-VI of the SAR includes a description of the rod bolts (M16) used for the bodies of the inner and outer receptacles. They are fabricated from chromium-molybdenum steel (JIS SCM435) and are identical for both receptacles. In addition, Table II-A-3 of the SAR includes the mechanical properties of the chromium-molybdenum steel.

11 The lids of the outer and inner receptacles are joined to the body of the outer and inner receptacles by twenty and sixteen rod bolts, respectively. Chapter II-A, Appendix 2, of the SAR includes the low-temperature characteristics of chromium-molybdenum steel. The chromium-molybdenum steel when treated by quenching and annealing obtains a high mechanical strength and keeps its tensile and impact characteristics at low temperatures described in Figures II-A.App2-3 (tensile characteristics) and II-A.App2-4 (impact characteristics) of the SAR.

The staff reviewed the temperature-dependent mechanical properties of the chromium-molybdenum used in the applicants mechanical calculation/testing and finds that the mechanical, corrosion, and fracture toughness properties are consistent with those found in technical literature. Therefore, the staff finds the chromium-molybdenum material to be acceptable for use in the design and construction of the Model No. GP-01 package.

3.5.3 Shock Absorbers Chapter I-C-VIII of the SAR describes the Model No. GP-01 ancillary elements. There are two urethane rubber guides that are provided on each of the four faces of the body of the inner receptacle to protect the aluminum honeycomb (alloy 5052, 5.7 PCF, 3/16-inch cell size, 0.002 nominal foil gauge, non-perforated) shock absorbers from damage that could be caused during loading into an outer receptacle. In addition, a nylon of high sliding capability is attached to the tip of the guide. Twelve anti-vibration rubber (JIS K 6386 Type CR chloroprene-rubber) plates are applied to the upper surface of the aluminum honeycomb elements on the bottom of the body of the outer receptacle to attenuate potential component of vibration during transport. The staff reviewed the temperature-dependent mechanical properties of aluminum alloy used in the applicants mechanical calculations/testing and finds that the properties are consistent with those in the technical literature (e.g., ASME Code Section II). Therefore, the staff finds the shock absorber materials to be acceptable for use in the design and construction of the Model No. GP-01 package.

3.6 Evaluation Findings

Based on a review of the statements and representations contained in the application, the staff concludes that the materials and evaluations have been adequately described, and the Model No. GP-01 package has adequate materials performance to meet the requirements of the IAEA SSR-6, 2012 Edition.

4.0 THERMAL EVALUATION The purpose of the thermal review is to verify that the GP-01 package design satisfies the thermal performance requirements in the 2012 Edition of the IAEA SSR-6, 2012 Edition. The staff applied the standards in SSR-6 for its review of the GP-01 package. The paragraphs from SSR-6 relevant to the thermal review were provided by the applicant in the form of a cross reference table [Cross Reference Table Between the Regulation Paragraphs and The Safety Analysis Report Paragraphs, dated March 11, 2020, (Document No. NFK-MPCF-2003002)]

(DOT, 2020a). The information from that document relevant to the thermal review is in the table reproduced below.

12 SSR-6 Paragraph SAR Section/Chapter Discussion 616 Chapter II-B.4 The design takes into account ambient temperatures and pressures for routine conditions of transport.

639 Chapter II - B.3 The brittle temperature for the silicone rubber provided at the containment boundary of the package is lower than -50°C.

Thus, the material will not suffer cracking or fracture in an environment of -40°C. The same silicone rubber can resist a temperature of 180°C. Therefore, it will not suffer cracking or fracture in an environment of 70°C.

679 Chapter II -

B.4.3 and B.4.2 The package is constructed with metallic materials, mainly stainless steels as structural elements. The brittle temperature for the silicone rubber provided at the containment boundary of the package is lower than -50°C.

These materials will not suffer cleft or fracture in an environment of -40°C. These metallic materials and the silicone rubber which resists a temperature of 180°C will not suffer cleft or fracture in an environment of 38°C. Thus, the package will not suffer cleft or fracture at temperatures ranging from -40°C to 38°C.

728 Chapter II - B.5 The package is subjected to the regulatory fire conditions (test performed).

A summary of the staffs review of the thermal characteristics and performance of the GP-01 package is discussed in the following sections.

4.1 Description of the Thermal Design The applicants approach to the thermal design and evaluation of this package, as described in the SAR, was three-fold:

1)

Prototype. A prototype with representative contents (dummy fuel pellets fabricated from lead) was subjected to a drop and thermal test (described below) in order to obtain representative component temperatures. The thermal test and results are discussed in Appendix 1 to Chapter II-B, Results of Prototype Thermal Test, of the SAR.

2)

Thermal model. A thermal model of the package was developed (in the ANSYS FEA code) and the fire test conditions were simulated in order to demonstrate that the model provides results in line with the experimental results. These results are discussed Appendix 2 to Chapter II-B, Results of Thermal Model Analysis for Integrating Thermal Test Results, of the SAR.

3)

NCT and ACT analyses. This model was then utilized for both NCT and ACT analyses of the package applying the appropriate boundary conditions as described in the governing requirements of the national Competent Authority which are equivalent to the requirements provided in IAEA SSR-6, 2012 Edition.

The development of the model is described in Section B.4.1 for NCT and Section B.5.1 for ACT while the highest temperatures obtained from the thermal analysis model are described in Section B.4.2. for NCT and Section B.5.3. for ACT.

13 The staff reviewed the following information related to the thermal evaluation for the GP-01 package:

1) package material thermal properties,

2) the descriptions of the physical testing of the package,

3) the assumptions used in the thermal analyses,

4) the descriptions of the thermal modeling of the package, and

5) the calculations related to the thermal models for NCT and ACT.

The staff reviewed the references above and determined that the applicant adequately described the thermal features of the Model No. GP-01 package and that this information is in alignment with the packages thermal evaluation.

4.2 Material Properties and Component Specifications The properties of the contents (materials contained in the inner receptacle) and the materials used in the fabrication of the GP-01 package are discussed in Sections II-B.2. and II-B.3. of the SAR, respectively.

4.2.1 Material Properties In Section II-B.2, Thermophysical Properties of Contents, of the SAR, the homogenized thermal properties used in the thermal analysis were calculated for the contents contained in the inner receptacle.

These values are presented in Table II-B-1 of the SAR (Page II-B-3). Properties of other package components, such as aluminum honeycomb, insulating material, and various types of rubber used in the package are used in the subsequent thermal models of the GP-01 package.

These properties are repeated in Appendix 2 (Tables II-B. App 2-3 to II-B. App 2-6 of the SAR).

4.2.2 Component Specifications Section II-B.3 includes a description of the characteristics of the package O-ring on the inner receptacle flange.

4.3 General Considerations 4.3.1 Evaluation by Test As discussed in Appendix 2 to Chapter II-B and Section II-B.5.1 of the SAR, a prototype test unit of the GP-01 was subjected to a single 9 m drop test, followed by a thermal exposure in accordance with Paragraph 728 of SSR-6 (IAEA, 2012), which stipulates a 800°C external ambient environment for no less than 30 minutes. The prototype was instrumented with accelerometers for the drop test, which were then replaced with 8 thermocouples for the thermal test. Temperature indicator strips and temperature sensitive paint were also used to record temperatures on the interior of the specimen.

14 The drop orientation was selected to cause the maximum damage to the upper corner of the package, with the intent of creating an opening in the flange region to potentially open a path for hot gasses to enter the package during the thermal test. No such opening was produced by the drop test. Appendix 1 to Chapter II-A of the SAR includes the results of the prototype drop tests.

The prototype test unit used dummy fuel pellets made from lead and a plain stainless steel plate (as opposed to a borated steel plate) for the neutron absorber. Table II-B. App 1-1 of the SAR provides a list of modifications made to the prototype package design used for testing that were adopted for the production version of the package.

The thermal test was conducted in a furnace that was heated to an indicated temperature of 1,000°C at which time the furnace door was opened, and the specimen was introduced into the furnace and the door closed. The temperature of the furnace was re-established to 800°C, which took 9 minutes (min.), and the specimen was left in the furnace for 30 min. The package was removed from the furnace and allowed to cool overnight. Thermocouple readings were taken for the duration of the cooldown.

4.3.2 Evaluation by Analysis The applicant provided a detailed description of the development of the analytical thermal model and the applied boundary conditions for:

a) comparison to the thermal test results of the prototype unit described above (in Section 4 of Appendix 2 to Chapter II-B of the SAR),

b)

NCT (Section B.4.1 of the SAR), and c)

ACT (Section B.5.1 of the SAR).

Models of the package were constructed for each of the separate analyses listed above using the ANSYS finite element analysis (FEA) code, were based on a quarter symmetry of the package, included all modes of heat transfer (i.e., convection, radiation, and conduction), and had roughly 120,000 elements each. There were also differences in the boundary conditions between the three analyses.

Geometry for the models is also described in Sections B.4.1.1.(1) for NCT and B.5.1.1.(1) for ACT. The model boundary conditions were specified in accordance with IAEA SSR-6, 2012 Edition, as described in Sections B.4.1.1.(2) for NCT and B.5.1.1.(2) for ACT. The ACT model included a representation of the damage to the exterior of the package body that was produced during the drop test. Summaries of the boundary conditions applied to the respective models are provided in Tables II-B-6 for NCT, II-B-8 for ACT, and II-B. App 2-2 of the SAR for the model compared to the thermal test results.

The results obtained from the models described above are provided below and summaries of the analyses results are provided in Tables II-B-7 for NCT, II-B-11 of the SAR for HAC, and II-B.App2-8 of the SAR where the maximum temperatures recorded during the thermal test are compared to those obtained in the corresponding package model.

The applicant used the results of these analyses as part of their demonstration that, for the content requested and previously authorized by the Japanese competent authority under

15 Certificate J/2009/AF-96, Revision1, (DOT, 2020a) the GP-01 package meets the thermal requirements in the IAEA SSR-6, 2012 Edition.

4.3.3 Contents Decay Heat As the proposed contents do not generate any appreciable decay heat, the decay heat of the contents was ignored for the thermal review.

4.3.4 Summary Tables of Temperatures Temperatures below are reported directly from the applicants SAR.

16 4.3.5 Margins of Safety As indicated above and further expanded upon in the applicants SAR, all package component temperatures fall inside the range of temperatures of concern for the various components examined. Therefore, the package would be able to meet the requirements of the IAEA SSR-6, 2012 Edition, for thermal performance and demonstrates a reasonable margin of safety for the proposed contents.

4.4 Thermal Evaluation under Normal Conditions of Transport 4.4.1 Heat and Cold The applicant examined the thermal performance of the package under the conditions stated for NCT in Paragraph 639 of the IAEA SSR-6, 2012 Edition, in Sections B.4.2 of the SAR for the Heat condition that includes solar heating, and B.4.3 of the SAR for Cold conditions. As indicated in the temperatures reported in Section 4.3.4 of this SER, the temperatures evaluated will not adversely affect the packaging.

4.4.2 Maximum Normal Operating Pressure In Section B.4.4. of the SAR, the applicant provides a simple calculation for the pressure difference between the interior and exterior of the inner receptacle under NCT (under solar heating and an inner receptacle of maximum temperature of 75°C) which shows a 27 kilopascals (kPa) gauge pressure.

4.4.3 Maximum Thermal Stresses In Section B.4.5. of the SAR, the applicant reports that the inner receptacle of the package will not experience any thermal stress due to differential thermal expansion.

4.5 Thermal Evaluation under Accident Conditions of Transport (ACT) 4.5.1 Initial Conditions Initial conditions for the ACT evaluation are defined in Paragraph 728 of the IAEA SSR-6, 2012 Edition, and discussed in Section B.5.(1) of the applicants SAR where it discusses the exposure of the package to a 38°C constant environmental temperature with a 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> off solar radiation cycle which was run until a constant surface temperature change pattern was observed. This is in accordance with the requirements in Paragraph 657 of the IAEA SSR-6, 2012 Edition.

17 4.5.2 Fire Test Conditions The fire test conditions were replicated in the thermal analysis, as discussed in Section B.5.(2) of the SAR and included an ambient environment of 800°C for 30 minutes. The ACT evaluation was conducted in accordance with the fire test conditions defined in Paragraph 728(a) of the IAEA SSR-6, 2012 Edition, with the conditions following the test as defined in Paragraph 728(b) of the IAEA SSR-6.

4.5.3 Maximum Temperatures and Pressure In Section B.5.4. of the SAR, the applicant provides a simple calculation for the pressure difference between the interior and exterior of the inner receptacle under ACT (for an inner receptacle maximum temperature of 170°C) which shows a 63 kPa gauge pressure, almost twice as much as for NCT conditions. The maximum temperatures for the package as a result of the ACT fire exposure are provided in Table II-B-11 of the SAR, which may be viewed Section 4.3.4 of this SER. None of the temperatures nor the pressures calculated exceed the limits of any of the materials used for fabrication of the packaging or the contents.

4.5.4 Maximum Thermal Stresses Although large temperature differences exist in the package body at the end of the thermal test, in Section B.5.5 of the SAR, the applicant claims that there will be no significant thermal stresses because the inner receptacle is not fixed to the outer receptacle of the package.

Similarly, it is not expected that differences in thermal expansion rates between package components will cause any significant additional stresses in the package.

4.5.5 Analyses Details As discussed in Section 4.3.2 of this SER, the applicant developed an ANSYS FEA model to evaluate the performance of the GP-01 package under the ACT fire test. Section B.5.1 of the SAR provides the details of the applicants model.

4.6 Evaluation Findings

Based on the review of the statements and representations in the application for the Model No.

GP-01 package, the staff concludes that the applicant adequately described and evaluated the thermal design of the Model No. GP-01 package. Therefore, the package meets the thermal requirements of IAEA SSR-6, 2012 Edition.

The staff concludes that the GP-01 package conforms to the requirements for type A packages loaded with fissile materials found in the regulations for the safe transport of radioactive material, IAEA SSR-6, 2012 Edition.

5.0 CONTAINMENT EVALUATION The purpose of the containment review was to verify that the GP-01 package design satisfies the requirements for the evaluation of the containment boundary as required in the IAEA SSR-6, 2012 Edition. The staff reviewed the application and confirmed that the GP-01 package containment system was appropriately evaluated for revalidation.

18 5.1 Description of the Containment System In the application, Chapter I, Description of Nuclear Fuel Package, the applicant provided a description of the containment boundary for the GP-01 package. Within the GP-01 packaging, the two major components are the inner receptacle and the outer receptacle, which protects the inner receptacle. The applicant also stated that the packaging is designed to store two assemblies of pellet storage boxes which contain pellets (minimum elements of nuclear fuel) of uranium oxides. The applicant, in Section II-C of the SAR, states that the containment boundary is formed by the inner receptacle. The body of the inner receptacle is open on the top to receive the intended contents. Once the contents are loaded into the inner receptacle, the lid, composed of a single 10-millimeter (10-mm) thick stainless-steel plate, is placed on the inner receptacle, along with an O-ring made of silicone rubber, and 16 rod bolts which are installed on the lid and tightened to securely close the lid. The actual containment boundary is formed by the O-ring seal that is secured under the load of the tightened bolts on the lid of the inner receptacle All the joints contributing to the containment boundary are finished with continuous welding.

The staff found that the design mentioned in the SAR meets the requirements of IAEA SSR-6 2012, Paragraph 650(b)(2).

In Section II-A, Structural Analysis, of Chapter II, Safety Analysis of Nuclear Fuel Package, of the SAR, the applicant describes the following:

1) the containment system,

2) the components of the inner receptacle,

3) how the inner receptacle is connected to the outer receptacle,

4) how to remove and tighten the rod bolts, and

5) how to transport the package.

The staff found that the design mentioned in the SAR meets the requirements of IAEA SSR 6 2012, Paragraph 650(b)(2).

In Sections A.5.3.2, A.8.2, and A.8.3 of Chapter II of the application, the applicant discusses the integrity of the containment boundary and the behavior of the package contents under NCT and ACT, respectively.

5.1.1 Temperatures (Highest and Lowest)

In Section B.4.2., Highest Temperatures, of Section II-B, Thermal Analysis (which was performed using the ANSYS FEA code), of Chapter II, the applicant provided a thermal analysis of the package and reported the highest temperatures that resulted. This evaluation was focused on the O-ring on the flange, which the applicant regarded as the most thermally vulnerable of the components of the inner receptacle and, therefore, the containment boundary of the package.

Figure II-B-7 of the application (i.e., SAR) presents the time-varying temperatures for NCT only.

Fig. II-B-8 shows the components evaluated. The temperature changes induced by solar

19 insolation practically attained equilibrium on the fifth day. The highest temperature for the O-ring was 68°C (154°F). This is lower than the O-rings maximum service temperature for normal service (180°C). The highest temperature (114°C) in the package was recorded in the insulator close to the outer receptacle lid center. Figure II-B-9 and Section II-B-10 of the SAR demonstrate the temperature distributions in the entire analytical model and in the O-ring and spacers, respectively, when the highest temperature was attained.

Stainless steel is the main material of fabrication for the transport packaging. Therefore, the temperature rise generated in the analysis will not adversely affect the packaging. The highest temperature (74.5°C) in the inner receptacle was generated near the lid center.

In Section B.4.3., Lowest Temperatures, of Section II-B, Thermal Analysis, of Chapter II, the applicant provided the lowest temperatures calculated for the package. The lowest ambient temperature was assumed to be -40°C. The contents of the package are pellets of unirradiated uranium oxides; therefore, the contents do not generate appreciable decay in the package.

Additionally, when solar radiation is neglected, the lowest temperature attained by the package was assumed to be the same as the applied ambient temperature (-40°C). Even if the temperature of the package is cooled down to -40°C, the materials of the packaging preserve their normal capabilities. The lowest normal service temperature for the O-ring is -50°C.

Therefore, the O-ring remains within its temperature limits at -40°C. Based on this description, the staff finds that the package meets the requirements of Paragraph 639 of the IAEA SSR-6, 2012 Edition.

5.1.2 Leaktight Analysis In Section II-C of the SAR, Leaktight Analysis, the applicant provided a discussion of the leaktight analysis performed for the GP-01 package. In this section of the SAR, the applicant provided a description of GP-01 transport packaging and the requested contents. The description includes a discussion about the use of the O-ring to seal the flange surface and the contents of the transport packaging. The SAR states that, under NCT, the leaktightness of the inner receptacle is maintained by the lid firmly joined to the body by means of rod bolts. In addition, an evaluation of the package for leaktightness [not as defined in ANSI N 14.5 (ANSI, 2014) regarding the performance of a leakage test done on a containment boundary] of the inner receptacle under NCT is provided in Section A.5.7, Summary of Results and Evaluation.

Regarding ACT, the SAR stated that packages configured with the Model No. GP-01 packaging are classified as type N" fissile transport packages and do not need to meet the regulatory requirements for leaktightness under ACT. The firm connection of the lid to the body of the inner receptacle, maintained by the rod bolts, contributes to containing pellet storage box assemblies in the inner receptacle.

5.1.3 Normal Conditions of Transport In Section C.3 of the SAR, Normal Conditions of Transport, the applicant states that, as described in Section A.5.7 of the SAR, Summary of results and evaluation, the evaluation of the package under NCT revealed that the leaktightness of the inner receptacle keeps radioactive materials/substances within its containment system and prevents them from leaking out.

20 The applicant describes its thermal test where:

a)

The O-ring temperature remains below the service temperature; b)

The applicant describes the results of the water spray test which will not affect the leaktightness of the package; c)

The applicant determined that the free drop test did not affect the leaktightness of the containment boundary of the package; d)

The stacking test was described, and it was demonstrated that the leaktightness of the inner receptacle would not be affected by the stacking test. The applicant discusses the penetration tests; and e)

The test rod for the puncture (drop) test itself would not penetrate the outer plates of the outer receptacle.

In Section C.3.1. of the SAR, the applicant concluded that the inner receptacle maintained its leaktightness under NCT and that no leakage of radioactive materials from the inner receptacle will occur. In Section C.3.2 of the SAR, the applicant discussed that if any rise of pressure occurs in the containment system, it will not affect the integrity of the inner receptacle. In Section C.3.3 of the SAR, the applicant states that the package/packaging contains no coolant, thus, no contamination of coolant will take place.

Based on this description, the staff finds that the package meets the containment requirements of IAEA SSR-6, 2012 Edition, Paragraph 653.

5.1.4 Accident Conditions of Transport (ACT)

The applicant provides a discussion of the analysis performed on the GP-01 packaging for ACT in Section C.4 of the SAR. The applicants tests and analyses have shown that the contents will be maintained in the package, and that none of the contents will leak from the package under ACT. The applicant discussed the following:

a)

After the first set of drop tests (Drop I Tests in Section C.4 of the SAR), no cleft or hole was produced in the joints on the flange, so that any exposure of the inner receptacle was prevented, and the radioactive contents will not leak from the storage boxes; b)

In the second set of drop tests (Drop II Tests in Section C.4 of the SAR), the applicant stated that several dents were produced on the external surface of the outer receptacle, but no penetration or crack/cleft or hole was generated on its external surfaces; c)

In the Thermal Test in Section C.4 of the SAR, the applicant discusses the results of the thermal test. The temperature of the O-ring for the inner receptacle may reach 170°C, which is below the maximum service temperature of the material of the O-ring, which is silicone rubber; and d)

In the Water Immersion Test in Section C.4 of the SAR, the applicant described the water immersion test.

21 In Sections C.4.1 and C.4.2 of the SAR, the applicant provided discussions on fission product gases and leakage of radioactive material, respectively. The pellets of unirradiated uranium oxides to be contained in the inner receptacle would not generate any fission product gas. With regards to the leakage of radioactive material, the contents would remain sealed (confined) in the inner receptacle.

Based on the description, the staff found that the package meets the requirements of IAEA SSR-6, 2012 Edition, Paragraph 503(c).

5.1.5 Inspections and Maintenance In Section B.0, Maintenance of Valves and Gaskets of Containment System, of Chapter IV, Handling and Maintenance of Nuclear Fuel Packages, of the SAR the applicant stated that whenever periodic voluntary inspections (at least once a year) are carried out, visual inspections should be carried out to verify that the O-ring on the inner receptacle flange does not present significant deterioration or partial loss of the receptacle flanges thickness that might affect its leak tightness.

In Section B.0 of the SAR, the applicant stated that these periodic and voluntary inspections may be omitted for packagings that are expected/planned to be stored without being used for at least one year in the same location; however, shortly before use for transport, at the end of long-term storage, such packaging should be subjected to the above visual inspections.

Sections B.1. Visual Inspections, B.7. Lifting Inspections, and B.10. Maintenance of Valves and Gaskets of Containment System, of the SAR include a description of these inspections.

In Section B.1, Storage of Packaging, of the SAR, the applicant stated that packagings that are expected/planned to be stored without being used for at least one year in the same location should be stored indoors.

Based on the description provided in the application, the staff found that the package meets the requirements of IAEA SSR-6, Paragraph 503(c).

5.2 Evaluation Findings

Based on review of the statements and representations in the Model No. GP-01 package application and its supplements, the staff concludes that the applicant adequately described and evaluated the containment system for the GP-01 package and that the package meets the containment requirements of the IAEA SSR-6, 2012 Edition. The staff recommends revalidation of the Japanese Certificate of Approval No. J/2009/AF-96, Revision 1 (DOT, 2020a).

6.0 CRITICALITY EVALUATION

The objectives of the criticality evaluation are the following:

1) confirming that the package will remain subcritical under routine conditions of transport, NCT, and ACT;

2) ensuring that arrays of packages under normal and accident conditions will remain subcritical; and

22

3) ensuring that the appropriate criticality controls for the package and for limiting the size of package arrays (the criticality safety index) are specified.

The staff performed its review to confirm compliance with the requirements relevant to criticality safety in the IAEA SSR-6, 2012 Edition. As a resource, the staff consulted the guidance in NUREG-1609, Standard Review Plan for Transport Packages for Radioactive Material, to inform the review.

6.1 Description of Criticality Design The staff reviewed the package description in Chapter I, Description of Nuclear Fuel Package, in the design drawings, and in Chapter II-E, Criticality Analysis, of the SAR. Section 1.1.1 of this SER includes a description of the package, including design aspects relevant to the criticality design and criticality analysis. The applicants criticality analysis does not include the outer containers legs and the recesses in the outer containers lid (nor the fusible plugs in the outer surfaces nor the bolts, though the bolts function of maintaining the lid in place is credited),

so they are not part of the criticality design. While the applicants criticality analysis does not include the packages rubber, insulation, and aluminum components materials, the applicants analysis does credit the space they occupy. The criticality analysis also ignores the rubber component materials and bolts of the inner container and its lid but credits the positioning of the pellet box assemblies in the containers cavity and the maintaining of the lids position that these components respectively provide.

As stated in Section 1.1.1.3 of this SER, the pellet box assemblies include rubber and steel components to maintain the pellet boxes vertical and horizontal position between the partitions and within the box assemblies. The applicants criticality analysis credits these components positioning function but neglects their materials as well as neglects the box assemblies posts.

The analysis also neglects the pellet boxes themselves except for their function of maintaining the fuel pellets within a fixed geometric volume. The applicant did not provide any specifications for the pellet boxes except for a description of their materials. The staff finds this to be acceptable because the applicants criticality analysis defines the dimensions that are set by the steel and rubber blocks that confine the pellet boxes positions within the box assemblies. The applicant described that each pellet box may be sealed in a separate plastic bag. This is an important consideration as it can lead to preferential flooding conditions within the package.

Also, since only one pellet box assembly type can be loaded into a package, the applicants criticality analysis includes calculations for packages containing two A assemblies and calculations for packages containing two B assemblies.

In its review of the application, including the drawings, the staff finds that the application contains sufficient information to describe the packaging components with respect to the criticality design and to support the criticality analysis. This information includes materials and dimensions, including tolerances. The staff also finds that the description of the design in the criticality analysis chapter is consistent with the information in the rest of the application.

The applicant performed a criticality analysis of the package using the SCALE code system, as described and evaluated later in this section of the SER. The applicant provided a table of maximum k-effective values that indicate the package k-effective will not exceed the applicant-determined upper subcritical limit and so the package will remain subcritical under all routine, normal, and accident conditions. The results in this table are for package arrays for routine conditions (an undamaged package) and accident conditions (a damaged package). The

23 applicant used the array results to bound the evaluation of a package in isolation for the requirements in Paragraph 682 of SSR-6. Since the arrays are infinite and because the applicant combined the damage from the normal condition tests with the damage from the accident conditions tests in the criticality analysis, the applicant used the accident conditions array to bound the normal conditions array. The applicant analyzed infinite arrays of packages; thus, following the requirements in Paragraphs 684 through 686 of SSR-6, the criticality safety index is 0.0.

6.2 Package Contents The package contents are as described in Section 1.1.2 of this SER. The staff reviewed the applications description of the contents and finds the description is clear and consistent in the application and provides the needed information for evaluating the criticality design of the package.

6.3 Considerations for Criticality Evaluations 6.3.1 General Model Configuration Considerations The requirements in SSR-6 (IAEA, 2012) include requirements for a package in isolation and package arrays. The requirements for packages in isolation must be subcritical for routine transport conditions, normal conditions, and accident conditions (see Paragraph 682 of IAEA SSR-6, 2012 Edition). The arrays are for packages under normal conditions and packages under accident conditions (see Paragraphs 684 and 685 of IAEA SSR-6, 2012 Edition). For packages that can be transported by air, a package in isolation must be subcritical under the Type C package tests as well (see Paragraph 683 of IAEA SSR-6, 2012 Edition).

The applicant did not develop models that were specifically for packages in isolation. Instead, the applicant developed models of infinite arrays of packages, basing that approach on the reasoning that the arrays would be bounding for single packages (with full reflection). The staff finds this approach to be acceptable because the packages, as modeled, result in significant interaction of neutrons between packages, which results in increased reactivity. In addition, the applicants array models adequately address each of the conditions that are to be evaluated for a package in isolation per Paragraph 682 of SSR-6 (IAEA, 2012).

Therefore, to address routine conditions, the applicant developed an infinite array of undamaged packages. This means the packages in the model use the as-designed package configuration. The applicant did not specifically develop a model for packages with impacts from the NCT tests. Instead, the applicant developed a model for an array of damaged packages that includes the damage from both the normal conditions tests and the accident conditions tests. The staff finds this approach to be acceptable since the array is infinite and the combination of damage from both sets of tests would be bounding for an infinite array of packages under NCT.

For both models, the applicant included only the steel and borated steel components of the packaging and the fuel material of the pellets. In its review of the model and the design drawings, the staff found that the included components were modeled at their nominal dimensions except for the borated steel plates in the inner container and the pellet box assemblies. The borated steel plates were modeled at their minimum dimensions, including for the thickness, width and height of each plate. All these components are relatively thin (the

24 thickest component is 10 mm) with quite small tolerances. Thus, the staff expects that the impacts of the tolerances of these components to be small.

The staff considered the tolerances in its review, including its confirmatory analyses. Based on these considerations, including the confirmatory analysis, the staff finds the selected dimensions in the models to be acceptable. The staff also evaluated the applicants choice to not include some of the other packaging components, such as the rubber components which could contribute to moderation between pellet boxes, between pellet box assemblies, and between the contents of adjacent packages. The applicant had provided a sensitivity analysis that indicated maximum water moderator density, for moderator in the fuel pellet zone resulted in maximum package array reactivity. Since the rubber components have a slightly higher density than water, the staff considered these components may result in increased reactivity. However, in its confirmatory analysis, the staff found that even small amounts of the rubber material between pellets boxes with optimum water moderator and small amounts of rubber material around the stack of pellet boxes reduced package array reactivity. Thus, the staff finds it acceptable to neglect the rubber materials in the model. Other materials like the aluminum components would simply act as absorber materials as the package was analyzed; thus, the staff finds that neglecting these materials is also acceptable.

Additionally, because each pellet box can be sealed individually within plastic bags, the staff finds that preferential flooding is a condition that the models and analysis needed to address.

The applicants models for both the undamaged array and the damaged array only include flooding within the pellet boxes. The applicants analysis left the remainder of the package cavities void. In this configuration, the applicant determined that full density water is the most reactive moderator density. The applicant also performed sensitivity studies for water of varying densities also being in the insulation area of the outer container as well as in the outer containers cavity around the inner container. Adding water to these areas of the package caused reactivity to decrease. The staff also performed some calculations that confirmed the applicants selected moderation configuration is the most reactive. Therefore, based on the applicants sensitivity study and the staffs confirmatory analysis, the staff finds that the applicant adequately addressed preferential flooding of the package and used the most reactive moderator, or flooding, conditions for both package arrays. The staff also finds that the modeled moderation conditions meet or bound the conditions required in Paragraphs 680, 684, and 685 of SSR-6 (IAEA, 2012) regarding water in the package.

The applicants model for the pellet box assemblies assumes the wall of the assemblies to be a solid item. However, the design drawings show multiple significantly large openings in the walls. With moderation present in the package, even when only in the pellet box regions, steel is a relatively good neutron absorber. Therefore, the staff finds that modeling the assembly wall as solid is non-conservative. In its independent calculations, the staff identified a small but noticeable increase in k-effective when modeling the box assemblies walls as void. However, the staff recognizes that the other components (e.g., the steel of the pellet boxes) that the applicants model neglects would result in a compensating reduction in reactivity were they included in the model. Thus, the staff finds the applicants model regarding the pellet box assembly walls to be acceptable.

In its review of the SAR, the staff determined that the applicant did not evaluate the package for air transport conditions. The applicant also confirmed that the request for revalidation does not include air transport. Therefore, the model configurations do not consider impacts from Type C package tests. Since the package is not analyzed for air transport, the staff recommends that the package revalidation be conditioned to preclude transport by air.

25 6.3.2 Undamaged Package Array Model For the undamaged package array, the applicant accounted for the dimensions of some of the packaging components that the applicant did not include in the model (i.e., the applicant credited the components dimensions but not their materials). For example, the applicant did not model the insulation and frames between the outer receptacles inner and outer shell but did model the spacing that they ensure between the shells. The applicant did not model the aluminum honeycomb components and the rubber anti-vibration pads but did model their effect on the position of the inner receptacle. The applicant made similar choices for positioning the pellet box assemblies in the inner container. The result is that the applicants model positioned everything in the nominal, as-designed configuration. The applicant also indicated that shifting of the pellet box assemblies and inner container from their nominal positions, would not have a noticeable influence on reactivity. This statement is supported by the applicants results for various sensitivity cases as well as the relatively small difference in k-effective between the undamaged and damaged array cases, in which, along with other differences, the pellet box assemblies and inner container positions are shifted more significantly than they could be in the undamaged package. Based on consideration of these results and confirmatory analyses, the staff finds the undamaged model to be acceptable.

The applicant modeled the pellets as long continuous cylinders of lengths equivalent to the short side of a pellet box. The applicant arranged the cylinders in rows just like the pellets would be in the trays in a pellet box, with enough rows at an optimized pitch to ensure the pellet box is loaded with the maximum amount of fuel allowed per box. Since the package is intended to transport pellets of any fuel type, the applicant did a sensitivity study versus pellet diameter for a range of diameters from 8 to 10 mm. The number of pellet cylinders was varied to keep the mass in the pellet box constant and the pellets pitch was varied. Based on the trends in k-effective with pellet diameter and the staffs knowledge of commercial pressurized water reactor (PWR) and boiler water reactor (BWR) fuel, the staff finds the evaluated pellet diameter range is acceptable to sufficiently cover those fuel types, which are UO2-based. The applicants study showed that the 8-mm diameter was the most reactive pellet diameter.

The staff finds the analysis acceptable for the other two uranium oxide compounds based on the applicants statement that these oxides are only impurities in the UO2 pellets. However, even if contents included pellets made of these other uranium oxides, physical properties of these pellets (e.g., density) and staff confirmatory calculations indicate that pellets of these oxides are less reactive than UO2 pellets. Moreover, the staffs judgement is that pellets of these oxides would not behave differently under routine transport conditions than UO2 pellets, including vibration. Therefore, the staff finds the criticality analysis for routine transport conditions is acceptable to cover all the package contents.

6.3.3 Damaged Package Array Model The applicant did not create a model for packages with damage from the NCT tests. Instead, the applicant created a damaged package array model that accounts for the combined damaged from both the NCT tests and the ACT tests. The applicant summarized the types of damage resulting from these tests in Section E.2.2 of the application. The staff reviewed the description of these damages and confirmed with the structural and thermal reviewers the descriptions of the damage that resulted from the package tests, descriptions of which are included in the structural and thermal sections of the application. The staff also confirmed with

26 these reviewers that the testing performed by the applicant results in the greatest damage to the package.

The applicant stated that it increased the damage represented in the damaged package model to be bounding of the damage observed in the NCT and ACT tests. This includes reductions of the width and length of the package by about double the measured reductions and a slightly larger reduction of the package height versus the measured reduction from the package tests.

The staff finds that model assumptions that introduce greater damage to the package are acceptable as they reduce the distance between moderated fissile materials in adjacent packages, which increases reactivity. The applicants model maintains the distance between the inner and outer shell of the outer container. In the applicants model, the distance between the inner surface of the aluminum honeycomb components and the outer surface of the outer shell is reduced slightly more than what was measured in the package tests, which is conservative. This includes complete loss of the aluminum honeycomb in the base of the outer containers cavity.

The borated steel plates on the inner containers walls remain attached in place. However, the borated steel plates that are the partition in the containers cavity are free to move from their designed position. The applicant did not model any other damage for the inner container. The applicant also considered that the pellet box assemblies move. The applicants model places the inner container in a lower corner of the outer containers cavity, with the inner container resting on the inner shell of the outer containers base. The applicants model also shifts the pellet box assemblies along with the borated steel partition plates to the matching corner of the inner container. The partition is still between the two pellet box assemblies. The applicant does not assume any other damage for the pellet box assemblies.

Since the applicant considered pellet fracture to be possible under the package test conditions, the applicant did a sensitivity study in which the pellet diameters were varied from 0 to 10 mm, the 0 mm case being a homogeneous mixture of fuel and moderator. The applicant varied pellet diameters and pitches while keeping the amount of UO2 at the maximum allowed per pellet box. The applicant also varied the stacks of rows or pellets as part of determining the most reactive configuration. The applicants analysis indicated that pellets with a diameter of about 4.5 mm in 5 layers of pellet rows with optimal spacing vertically and horizontally is the most reactive configuration. Therefore, the applicants most reactive damaged package array case uses this configuration in the pellet boxes together with the shifting of packaging components and dimensional changes described above.

In reviewing the applicants models, the staff compared the modeled damage to the damage resulting from all the different drop tests for both the NCT tests [i.e., the tests specified in SSR-6 Paragraph 684(b) of IAEA SSR-6, 2012 Edition] and the ACT tests [i.e., the tests specified in SSR-6 Paragraph 685(b) of IAEA SSR-6, 2012 Edition] for both prototype packages that were tested. The staff found that for much of the described damage, the criticality model is bounding.

The staff also identified other damage which was not bounded by the damage in the criticality model. This damage, like the damage which the applicant considered, did not occur uniformly over the packages outer container but was localized. The staff, as part of its confirmatory calculations, evaluated the impacts of adding this extra damage, applying it conservatively over the effected surfaces of the outer container. This extra damage resulted in a more compact package. The staff also identified discussions of somewhat limited damage to the neutron absorber plates between the pellet box assemblies and included that in its confirmatory models.

The inclusion of these effects did slightly increase the reactivity of the damaged package array.

However, the staff recognizes that other conservatisms in the applicants model, even with only

27 small relaxations (e.g., credit for some of the insulation material in the outer containers wall) would be sufficient to compensate for the further crush damage and damage to the neutron absorber plates that the staff evaluated.

The staff reviewed the models with regard to the shifting of the inner container, the pellet box assemblies, and the neutron absorber plates between the pellet box assemblies. The staff finds this shifting to be either consistent with or bounding for the shifting (and possibility of shifting) that the measured damage in the package interior would allow. With reflective boundary conditions to simulate an infinite array of damaged packages, this places the fuel contents of eight adjacent packages as close to each other as possible. The staff noticed that the package damage descriptions relevant to the amount of possible movement of the inner container within the outer containers cavity seemed to be inconsistent, with some of the description indicating the possibility that the inner container could be closer to the outer containers inner walls than it is in the applicants model. For this reason, the staff also performed a confirmatory calculation to evaluate the impact of placing the inner container closer to the outer containers inner walls.

The staffs calculation indicated that the effect was negligible. Therefore, the staff finds the applicants model configuration in this respect to be acceptable.

The staff also identified that, for at least one of the prototypes, there was damage to the box assemblies and slight damage to the pellet boxes. The applicants models do not include damage to these components and uses the undamaged configurations. The kinds of damage that were seen in the package tests had the potential to impact the moderation conditions of the pellets. The applicants analysis models use the moderation configuration that maximizes reactivity. Based on that configuration and the staffs confirmatory calculations related to identification of maximum reactivity moderation conditions, the staff finds that the damage incurred in these components would result in moderation conditions that would reduce reactivity.

Thus, the staff finds the applicants model with regard to the damage to the pellet boxes and pellet box assemblies (i.e., modeling them as undamaged) to be acceptable.

The staff considered whether modeling the fractured pellets as cylinders of different diameters was conservative since the fractured pellet pieces can take any shape. In its confirmatory calculations, the staff selected one diameter and used that as the diameter of a sphere, which the staff then surrounded with moderator in a dodecahedron. Keeping the mass of fuel at the maximum allowed per pellet box, the staff modeled the fuel fragments in an array of spheres.

This resulted in a k-effective that was less than nearly all of the k-effective results of the applicants sensitivity study. Therefore, the staff finds that the applicants pellet sensitivity study is acceptable to determine maximum reactivity for fuel fragments in the damaged package array.

The staff finds the analysis acceptable for the other two uranium oxide compounds based on the same reasons the staff noted above for finding the undamaged package array analysis to be acceptable. In addition, for the damaged packages, the analysis considers a wide range of pellet diameters, including a 0 mm diameter and homogenized mixture of fuel and moderator.

The staff finds that this would address any behavior of these pellets in response to the mechanical impacts of normal conditions and accident conditions, since the damaged array model is for both normal and accident conditions. This further supports the staffs finding the damaged package analysis acceptable for pellets of these two additional uranium oxides.

28 6.3.4 Material Properties Application Section E.3.2 includes a description of the properties of the materials the applicant included in the model. The models include UO2 at theoretical density and 5 wt. % enrichment (the maximum allowed), stainless steel, borated stainless steel with 1 wt. % boron, and water.

The staff evaluated the densities of the materials and finds the steel densities to be acceptable.

While the actual borated stainless steel has 1 wt. % of boron, the modeled borated stainless steel only credits 75 % of that, or 0.75 wt. % of boron. The staff finds this acceptable as it is consistent with how borated materials are considered in criticality analyses for domestic packages. Based on these properties, the applicant determined the atom densities of each of the materials constituent elements and nuclides in the model. These atom densities are shown in Table II-E-1 of the SAR. The staff reviewed these atom densities and finds that while the table in the application shows incorrect values for the respective constituent elements and nuclides in the materials except for the uranium nuclides, this is the result of an error that shifted the atom densities by one row. The applicant corrected this error in its response to staff questions (DOT, 2021). The staff confirmed that the corrected table shows correct atom densities of all elements and nuclides shown in the table. The applicant states that the models use the appropriate standard composition libraries that are available in the SCALE code system.

Based on this review, the staff finds the materials and their properties for the models to be acceptable.

6.3.5 Analysis Methods and Nuclear Data section The applicant performed its criticality analysis using the Keno V. a code in Version 5 of the SCALE code system. This is a somewhat older version of the SCALE code system. However, it was developed specifically for the purposes of performing criticality analyses, including of fissile material packages. The code has undergone verification and validation to demonstrate its suitability for use and the staff has accepted the use of this version of the code in previous applications. The staff determined that the current application is within the bounds of previously accepted uses of the code and so finds the use of the code in this application to be acceptable.

The applicant used the 44-group cross section library based on ENDF/B-V data. This cross-section library was established based on the spectrum for light water reactor fuel, which is the type of content of this package. The staff did identify that the pitch in one direction did not match the pitch in the other direction. For that reason, the staff considered that the cell data for specifying cell pitch for the pellets, which is needed for using multigroup cross sections libraries, may lead to some nonconservative results due to cross section processing being based on a uniform pitch in both directions across the cell cross sectional area. The staff did notice that the differences in the pitch in the two directions for optimized pitches was noticeable, though it was not very large (an 8% difference in one case).

As part of confirmatory calculations, the staff, which used SCALE 6.2.3 and the csas26 (Keno--VI) criticality sequence, evaluated the differences resulting from use of group cross sections and CE cross sections. For the multigroup cross section case (238-group ENDF/B-VII cross section library), the staff specified a pitch in the cell data that resulted in the equivalent cross section area as the actual pellet cell cross section in the model geometry. Comparing the k-effective result for that calculation with the result for the continuous energy cross sections from ENDF/B-VII.1 data, the staff identified only a very small difference. Although, the staffs group library differs from the one used by the applicant, the 44-group library is specifically for light water reactor fuel and the materials evaluated in package criticality model. Additionally, the cross section data for the materials in the model have been consistent and well-defined across

29 the various versions of ENDF/B data. Thus, the staff considers that a comparison with the k-effective results with use of the 44-group library and the results when the continuous energy cross section data is used will likewise show a very small effect on reactivity. Based on these considerations, the staff finds the applicants use of the 44-group cross sections with the related cross section processing to be acceptable.

6.3.6 Demonstration of Maximum Reactivity and Criticality Safety Index The staff reviewed the applicants analysis, including the sensitivity studies on various parameters. The studies include effects of pellet size and pitch and different package flooding configurations. Based on these evaluations, the applicant determined the configurations that result in maximum package reactivity. The applicant provided the maximum k-effective plus three times the calculational standard deviation for both an infinite array of undamaged packages and an infinite array of damaged packages. The results for these arrays include arrays of packages containing the two different types of pellet box assemblies. The arrays of packages with the Type A assemblies, which have the larger mass of fuel, are the most reactive. The damaged arrays are also the most reactive array types. These array k-effective values bound those for packages in isolation. The applicants maximum k-effective plus three times its standard deviation is 0.928, which is below the applicants restriction value (i.e., its subcritical limit) of 0.95.

The staff also conducted analyses to confirm that the applicant properly identified the most reactive configuration. This analysis included calculations that accounted for tolerances of all modeled packaging components, the effects of including rubber components materials, and shifts in package contents among others. Some of these calculations have been mentioned above. Based on a review of the applicants analysis and the staffs confirmatory calculations, the staff finds the applicant has demonstrated the maximum k-effective for the package. The staff also finds that the applicants CSI of 0.0 is appropriate since the arrays are infinite.

6.3.7 Confirmatory Calculations As noted already, the staff performed a variety of confirmatory calculations. The staff used the latest version of the SCALE code system, SCALE 6.2.3. The staff used the csas6 criticality sequence, which uses Keno-VI geometry. This allows greater flexibility in defining the geometry of the model (e.g., using dodecahedrons to create arrays of spherical pellet fragments). For calculations with multigroup cross sections, the staff used the 238-group cross section library from ENDF/B-VII data. For calculations with continuous energy cross sections, the staff used the ENDF/B-VII.1 cross section data. The staffs calculations for the applicants maximum reactivity cases resulted in k-effective values that were similar to, or bounded by, the applicants results.

6.4 Benchmark Evaluation The applicant performed a benchmark evaluation. Consistent with the practices in the applicants country, the applicant used the benchmark evaluation to show that the calculation method produces results with satisfactory accuracy but did not use it to determine a bias and bias uncertainty to apply to the package analysis.

The applicant analyzed 340 experiments within the category of the LEU-COMP-THERM (i.e.,

low enriched uranium, compounds, thermal neutron spectrum) experiments found in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. These

30 experiments include those that involved UO2 powder and U3O8 powder. The vast majority are experiments with UO2 pellet-rod lattice arrays.

In addition to the information in the application, the applicant provided some further information in response to the staffs questions. The information includes a qualitative evaluation for trends on calculated k-effective as a function of pellet diameter, enrichment, moderator-to-fuel volume fraction, pellet pitch, fuel density, and the energy of the average lethargy causing fission. The information includes figures for each parameter that shows the most reactive package configuration versus the experiments for which k-effective was calculated in the benchmark analysis. The figures show the packages most reactive case is within the benchmark analysiss area of applicability for every parameter.

In reviewing the applicants benchmark evaluation as described in the application and the applicants response to the staffs questions, the staff finds that there is very little information to identify which LEU-COMP-THERM experiments the applicant evaluated. The staff also identified that the applicants evaluation does not include evaluation of subsets of the experiments and the implications on bias and bias uncertainty and trends in the bias (e.g., evaluations without the powder experiments vs. evaluations that include the powder experiments). While the applicant states that it did not identify any trends in k-effective (and so in bias) relative to any parameter (e.g., enrichment, pellet diameter), the staff identified that at least some of the figures show trendlines that appear to have a clear trend (a non-zero slope) or otherwise indicate a clear trend over a portion of the data in the figure. However, since the applicant did not provide data about the evaluated experiments, the calculation results for those experiments, the trendlines equations, or trendlines R2 values, the staff cannot make a more definitive determination regarding potential trends in the data. Also, since the applicant did not include a trend on boron absorber concentrations, the staff finds that it is not clear how the data behave in relation to that parameter. This is a relevant parameter since the package has borated absorbers and many of the analyzed experiments likely include borated and other absorbers as well.

The staff notes that if the applicants average k (i.e., the difference between the experimental k-effective, taken to be unity, and the calculated k-effective for the experiments) plus the confidence coefficient times the standard deviation of that k were used as a bias and bias uncertainty, then applying them to the margin used to get the restriction value (the applicants subcritical limit) would result in an upper subcritical limit that is equal to the k-effective of the maximum reactivity case in the applicants analysis plus three times its standard deviation.

Based on a review of the information in the applicants response to the staffs questions (DOT, 2021), the staff estimated that the k at the point equal to the most reactive package case could be as large as 0.006. This estimate includes the staffs qualitative estimate of the impacts of neglecting the data from the powder experiments. Based on the data shown in the figures that identify those experiments, the staff estimates that at least for some of the parameters, the trendline would adjust lower than currently shown in the figures by a small amount. The staff could not estimate how neglecting those data would affect the standard deviation, but the staff considered the possible effect of the standard deviation of the average k increasing from 0.008 in the application to 0.010.

Using the above estimates, the staff estimated upper subcritical limits based on U.S. practices for benchmark evaluations. Using the applicants standard deviation on the average k, the staff calculated a limit of 0.9268. using the larger standard deviation, the staff calculated a limit of 0.9225. Per U.S. practices, the maximum package k-effective plus twice its standard deviation is compared against the limit. Based on the information in the application, this value

31 for the GP-01 package is 0.927. This value exceeds both upper subcritical limits, though only slightly for the first limit value and by about 0.45 % for the second limit. Thus, this process would indicate the most reactive package case (infinite array of damaged packages) may not be subcritical. However, the staff finds that the applicants package model includes some conservatisms that could be changed to credit more of the packaging components in a way that is still consistent with or bounding for the combined normal conditions and accident conditions impacts. It is the staffs judgement that such changes would result in a maximum k-effective plus twice its standard deviation that does not exceed either of the staffs estimated upper subcritical limits, though more packaging components would need to be credited to not exceed the lower of the staffs two estimated limits. Additionally, based on the package having been reviewed and approved by the designers countrys regulatory authority and the consistency of the staffs confirmatory analysis results compared to the applicants analysis results, the staff has reasonable assurance that the package, including an infinite array of damaged packages, will be subcritical even with a benchmark evaluation similar to what is done in the U.S.

6.5 Operations, Acceptance Tests and Maintenance Programs Related to Criticality Safety The staff reviewed the package operations descriptions in the application as relates to criticality safety. There are no features that require what the staff considers unusual operations. The operations descriptions include verifications of proper contents and packaging component and pellet box integrity. The staff finds the operations descriptions to be adequate to ensure subcriticality of the package during transport and to ensure that the package is used in a manner that is consistent with or bounded by the criticality analysis in compliance with Paragraph 502 of SSR-6 (IAEA, 2012).

The staff also reviewed the acceptance tests and maintenance programs descriptions with respect to ensuring criticality safety. The staff finds that the maintenance program includes adequate verifications of packaging components to ensure the criticality safety function of the package is maintained consistent with the package design. These verifications include confirmation of the positioning and condition of the borated stainless steel plates in the inner container. The staff finds that the acceptance tests also include the necessary checks to confirm the as-fabricated package meets the package design and so ensures criticality safety for the as-fabricated package in compliance with Paragraph 501(c) of SSR-6 (IAEA, 2012).

These checks include checks that the physical condition, dimensions, and chemical composition of the borated stainless steel plates in the inner package and in the box assemblies (i.e., the partition plates between pellet boxes in the box assemblies) meet the specifications in the package drawings and used in the criticality analysis. They also include neutron transmission tests of these borated stainless steel plates to confirm their neutron absorption efficacy.

6.6 Evaluation Findings

Based on a review of the information in the application, including the design drawings and design descriptions and the criticality analysis, and the staffs confirmatory calculations, the staff finds, with reasonable assurance, that the package satisfies the criticality safety requirements (e.g., Paragraphs 682 and 684 through 686) in SSR-6, the 2012 Edition (IAEA, 2012).

As described previously in this section of the SER, the applicant did not include air transport as part of the revalidation request nor did the applicant evaluate the package for air transport.

Therefore, the staff recommends that the revalidation of the packages certificate be conditioned to preclude air transport.

32 7.0 QUALITY ASSURANCE The purpose of the quality assurance (QA) review is to verify that the package design meets the requirements of the IAEA SSR-6, 2012 Edition. The staff reviewed the description of the QA program for the Model No. GP-01 package against the standards in the IAEA SSR-6, 2012 Edition.

7.1 Evaluation of the Quality Assurance Program The applicant developed and describes a QA program for activities associated with transportation packagings for nuclear fuel materials. Specifically, the principles of the management system to be applied during the design, manufacture, inspection, testing, maintenance, and use of the package are described in CHAPTER III - Basic Rules and Principles for Quality Management of the SAR. Further, the staff notes that the applicants description of the QA program is based on ISO 9001, Quality Management System and JEAG 4111, Management System Regulations for Nuclear Safety; therefore, it satisfies the requirement of IAEA SSR-6, 2012 Edition, Section 306, Management System.

The staff finds the QA program description acceptable, since it allows implementation of the associated QA program of the Model No. GP-01 transportation package.

7.2 Evaluation Findings

Based on review of the statements and representations in the application for the Model No.

GP-01 and, as discussed in this SER section, the staff has reasonable assurance that the GP-01 package meets the requirements in IAEA SSR-6, 2012 Edition. Therefore, the staff recommends revalidation of Japanese Competent Authority Certificate of Approval J/2009/AF-96, Revision 1 (DOT, 2020a).

8.0 REFERENCES

(ANSI, 2014)

American National Standards Institute (ANSI) N14.5, Radioactive Materials - Leakage Tests on Packages for Shipment, ANS, LaGrange Park, IL.

(IAEA, 2012)

International Atomic Energy Agency, IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, 2012 Edition, https://www-pub.iaea.org/MTCD/Publications/PDF/Pub1570_web.pdf.

(DOT, 2020a)

Boyle, Richard W., U.S. Department of Transportation (DOT), letter to Andrea Kock, U.S. Nuclear Regulatory Commission (NRC), May 11, 2020, ADAMS Package Accession No. ML20141L696.

(DOT, 2020b)

Boyle, Richard W., U.S. Department of Transportation (DOT), letter to Norma Garcia Santos, U.S. Nuclear Regulatory Commission (NRC),

August 5, 2020, ADAMS Package Accession No. ML20231A505.

33 (DOT, 2021)

Boyle, Richard W., U.S. Department of Transportation (DOT), letter to Norma García Santos, U.S. Nuclear Regulatory Commission (NRC),

January 12, 2021, ADAMS Package Accession No. ML21013A481.

(TNI, 2018)

TN International (TNI), Safety Analysis Report GP-01, NFK-MPC-1801024, January 2018, ADAMS Accession No. ML20247J319.

CONDITIONS The staff recommends the revalidation of Japanese Competent Authority Certificate of Approval J/2009/AF-96, Revision 1, for the Model No. GP-01 package, with the following additional condition:

Transport by air of the Model No. GP-01 is not authorized.

CONCLUSION Based on the statements and representations contained in the documents referenced above, and the conditions listed above, the staff concludes that the Model No. GP-01 package meets the requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012).

Issued with letter to R. Boyle, U.S. Department of Transportation, on March 1, 2021.