ML19213A037

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Letter to R. Boyle Certificate of Approval No. J/61/B(U)F-96, Revision 3, for the Model No. JCR-80Y-20T Packaging (Docket No. 71-3035) (W/Enclosure 1)
ML19213A037
Person / Time
Site: 07103035
Issue date: 08/01/2019
From: John Mckirgan
Spent Fuel Licensing Branch
To: Boyle R
US Dept of Transportation (DOT)
Garcia-Santos N
Shared Package
ML19213A035 List:
References
EPID L-2018-LLA-0219, JCR-80Y-20T
Download: ML19213A037 (16)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 1, 2019 Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 400 Seventh Street, S.W.

Washington, D.C. 20590

SUBJECT:

CERTIFICATE OF APPROVAL NO. J/61/B(U)F-96, REVISION 3, FOR THE MODEL NO. JCR-80Y-20T PACKAGING (DOCKET NO. 71-3035)

Dear Mr. Boyle:

This is in response to your letter dated August 8, 2018 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML18270A151), as supplemented on February 19, 2019 (ADAMS Package Accession No. ML19059A420) requesting our assistance in evaluating the Model No. JCR-80Y-20T package authorized by Japanese Certificate of Approval No. J/61/B(U)F-96, Revision 3. The certificate holder requested to add JRR-4 high enriched uranium fuel as authorized content of the Model No. JCR-80Y-20T.

Based upon our review, 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/61/B(U)F-96, Revision 3, for the Model No. JCR-80Y-20T package.

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

Sincerely,

/RA/

John McKirgan, Branch Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3035 EPID L-2018-LLA-0219

Enclosures:

1. Safety Evaluation Report
2. Certificate from Competent Authority

R. Boyle

SUBJECT:

CERTIFICATE OF APPROVAL NO. J/61/B(U)F-96, REVISION 3, FOR THE MODEL NO. JCR-80Y-20T PACKAGING (DOCKET NO. 71-3035),

DATE: August 1, 2019 DISTRIBUTION: SFM r/f, DMarcano This closes EPID L-2018-LLA-0219.

\\hqvwxalic01\XenAppUserProfiles\ngs\Desktop\JCR-20Y-2OT_Revision_3_SER.docx ADAMS Package Accession No.: ML19213A035 LTR & SER: ML19213A037 ENCL 2: ML19213A036 OFFICE: DSFM DSFM DSFM DSFM DSFM DSFM NAME: NGarcía SFigueroa CBajwa YKim DTarantino JSmith Santos by email by email by email by email by email DATE: 7/12/19 7/26/19 7/29/19 7/16/19 7/19/19 7/18/19 OFFICE: DSFM DSFM DSFM DSFM DSFM NAME: ASotomayor YDíaz TTate by MRahimi JMcKirgan by email Sanabria email by email by email DATE: 7/26/19 7/31/19 7/26/19 7/18/19 8/1/19 OFFICIAL RECORD COPY

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3035 Model No. JCR-80Y-20T Package Certificate of Approval No. J/61/B(U)F-96 Revision 3 Enclosure 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 ......................................................................................................................... 2 2.0 STRUCTURAL EVALUATION ....................................................................................... 2 2.1 Description of Structural Design and Analysis ................................................................ 2 2.2 Weights and Centers of Gravity ...................................................................................... 2 2.3 Evaluations ..................................................................................................................... 2 2.4 Evaluation Findings ........................................................................................................ 3 3.0 MATERIALS EVALUATION .......................................................................................... 3 4.0 THERMAL EVALUATION .............................................................................................. 4 4.1 Description of the Thermal Design ................................................................................. 4 4.2 Thermal Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions ..................................................................................................................................... 4 4.3 Evaluation Findings ........................................................................................................ 4 5.0 CONTAINMENT EVALUATION ..................................................................................... 4 6.0 SHIELDING EVALUATION ............................................................................................ 4 6.1 Shielding Design............................................................................................................. 5 6.2 Shielding Evaluation ....................................................................................................... 5 6.3 Evaluation Findings ........................................................................................................ 9 7.0 CRITICALITY SAFETY EVALUATION .......................................................................... 9 7.1 Description of Criticality Design ...................................................................................... 9 7.2 Spent Nuclear Fuel Contents ......................................................................................... 9 7.3 General Consideration for Criticality Safety .................................................................. 10 7.3.1 Evaluation of Single package ....................................................................................... 10 7.3.2 Evaluation Array of packages ....................................................................................... 10 7.3.3 Criticality Safety Index (CSI) ......................................................................................... 10 7.4 Demonstration of Maximum Reactivity ......................................................................... 10 7.5 Evaluation Findings ...................................................................................................... 11 8.0 CONDITIONS ............................................................................................................... 11

9.0 REFERENCES

............................................................................................................. 11 CONCLUSION ............................................................................................................................ 12 List of Tables Table 6.1: Applicable IAEA, SSR-6, 2012 Edition, Shielding Requirements to Evaluate the Adequacy of a Package Design [Type B(U) Package]. ................................................................ 5 Table 7.1: Applicable IAEA, SSR-6, 2012 Edition, Criticality Requirements to Evaluate the Adequacy of a Package Design [Type B(U) Package]. ................................................................ 9

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3035 Model No. JCR-80Y-20T Package Certificate of Approval No. J/61/B(U)F-96 Revision 3

SUMMARY

By letter dated August 8, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18270A163), as supplemented on February 19, 2019 ADAMS (Accession No. ML19059A412), the U.S. Department of Transportation (DOT) requested the review and recommendation from the U.S. Nuclear Regulatory Commission (NRC or staff) regarding the revalidation of the Japanese Certificate of Approval No. J/61/B(U)F-96, Revision 3. The certificate holder requested to add JRR-4 high enriched uranium (HEU) fuel as authorized content of the Model No. JCR-80Y-20T package (referred as Model No. JCR-80Y-20T). The Model No. JCR-80Y-20T is a Type B(U) package.

The NRC reviewed the information provided to the DOT by Edlow International Company (the applicant) in its application for the Model No. JCR-80Y-20T package and its supplements against the regulatory requirements of International Atomic Energy Agency (IAEA), Safety Standard Series, SSR-6, Regulations for the Safe Transport of Radioactive Material, 2012 Edition (IAEA, 2012a). Based on the statements and representations in the information provided by DOT and the applicant, the staff recommends the revalidation of Certificate of Approval No. J/61/B(U)F-96, Revision 3, Model No. JCR-80Y-20T package, for the contents included in Section 1.1.2, Contents, of this safety evaluation report (SER).

1.0 GENERAL INFORMATION The DOT provided a safety analysis report (SAR), Revision 3, for the JRC-80Y-2OT package and indicated that the NRC has previously reviewed the SAR, Revision 1, under Docket No.

71-3035 (ADAMS Accession No. ML061460373). The applicant proposes to include an additional content, "JRR-4 high enrichment instrumented type fuel (HEU), to the JRC-80Y-20T package. The SAR, Revision 2, was reviewed by the DOT.

1.1 Package Description 1.1.1 Packaging The main structures of the package consist of the shell, bottom plate, the lid, and the basket.

The package shell is a solid stainless-steel forging weighing approximately 23 metric tons (20.87 tons) when loaded. The package has a lid with double O-rings and it is secured by 16 bolts. The lid has a test port for leak testing and a vent port with a cover plate sealed with a double O-ring. The bottom shell has an installed drain port, which is also sealed by a port cover plate and double O-ring seals. The shell, lid, and bottom plate also serve as shielding. The applicant did not request changes to the packaging design.

2 1.1.2 Contents The applicant requested approval to add a new content, JRR-4 HEU, fuel for revalidation of the transportation package for import and export use. The staff reviewed Table 1, Specification of Radioactive Contents, of the SAR for JRR-4 HEU fuel (e.g., fuel type, burnup, decay heat, fuel material, etc.). The staff notes that the JRR-4 HEU uranium aluminum alloy fuel is composed of similar materials as previously approved fuels and the aluminum alloy cladding is identical to previously approved cladding materials in this package.

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

2.1 Description of Structural Design and Analysis The applicant used identical structural design criteria1 and analytical methods (i.e., LS-DYNA code, etc.) for the structural analysis of the proposed content (i.e., JRR-4 HEU) as the design criteria and analytical methods used for approved fuel types2 for transport in the Model No.

JRC-80Y-20T package. The applicant used combinations of closed-form solutions, hand calculations, and finite element analysis methods to evaluate the JRR-4 HEU.

2.2 Weights and Centers of Gravity The applicant calculated the weights and centers of gravity for the JRR-4 HEU in Section II-A-2 of the SAR. The values of the weight and the location of the centers of gravity for the JRR-4 HEU are provided in (II) Figure A.1 and (II) Table A.2 of the SAR. The staff notes that the values of the weight and the location of the centers of gravity of the JRR-4 HEU are identical to the values of the weight and the location of the centers of gravity for the JRR-4 LEU, which were previously reviewed and accepted in the SAR, Revision 1.

2.3 Evaluations The staff focused on reviewing Section II-A of the SAR, Revision 3, with emphasis on the hypothetical accident conditions for the JRC-80Y-2OT package including the JRR-4 HEU, since the drop analysis under the normal conditions of transport is bounded by the hypothetical accident conditions. The applicants drop analysis for the hypothetical accident conditions was performed using the LS-DYNA finite element code. For the 30 feet (ft.) drop, the staff observed that the highest stresses in the lid bolts occurred for the center of gravity over top corner drop.

This is consistent with the previous submittal of the JRC-80Y-20T package in the SAR, Revision

1. This is expected because the following characteristics of the package remained the same as in the SAR, Revision 1:

1 i.e., applied loads, allowable stresses, strain criteria, impact & energy absorption, vibration, failure modes, etc.

2 JRR-3 standard aluminide type, JRR-3 standard silicide type, and JRR-4 low enrichment (LE) silicide type

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a. overall dimensions,
b. center of gravity, and
c. weight and construction.

Therefore, the Model No. JRC-80Y-20T package with the proposed content (e.g., JRR-4 HEU) should experience the same g-loads as observed in the previously approved version (SAR, Revision 1).

In addition, the staff reviewed the results of the structural analysis provided in (II) Table A.20 of the SAR and found that the calculated factors of the safety (FS) are 2.0 and 9.0 for the fuel side plate and fuel plate, respectively, demonstrate that the package has the strength and safety margin to withstand hypothetical accident conditions as required in IAEA regulations (IAEA, 2012a). The staff also examined the calculated FS of various components of the package during impact and found that their FS is consistent with the results of the structural analysis in SAR, Rev. 1, which the staff previously reviewed and accepted. Therefore, the staff finds that the structural analysis and evaluation of the JRC-80Y-2OT package with the JRR-4 HEU are acceptable.

2.4 Evaluation Findings

Based on the review of the statements and representations contained in the application, the staff finds that the Model No. JRC-80Y-20T transportation package with the JRR-4 HEU meets the regulation requirements of IAEA SSR-6, 2012 Edition.

3.0 MATERIALS EVALUATION The staff reviewed the adequacy of the package materials of the Model No. JCR-80Y-20T against the IAEA SSR-6, 2012 Edition, (IAEA, 2012a) requirements related to the materials performance of the package, and the ability of the package design to meet such requirements.

The applicant did not request changes to the packaging design. The applicant requested approval to add a new content, JRR-4 HEU fuel, for revalidation of the transportation package for import and export use. The package is designed to meet the general use, function, and testing requirements specified in the IAEA Safety Standard Series document SSR-6 (IAEA, 2012a). The package is made of welded stainless-steel and consists of a canister body holding a fuel basket containing fuel elements.

The staff finds that the mechanical and heat transfer properties of the materials used in fabrication of the Model JRC-80Y-20T package are acceptable to transport JRR-4 HEU fuel.

The staff bases this finding on the JRR-4 HEU fuel characteristics and their effect on the packaging materials being bounded by previously approved fuels. The new fuel content does not expose the packaging materials to temperatures, radiation dose, or other environmental exposures that were not previously evaluated by the staff.

4 4.0 THERMAL EVALUATION The purpose of the thermal review is to verify that the changes to the package design meet the requirements for the thermal evaluation required under the IAEA SSR-6, 2012 Edition (IAEA, 2012a). The staff reviewed the thermal properties of the materials used for the Model No. JCR-80Y-20T package and the description of the thermal analysis against the standards in the IAEA SSR-6, 2012 Edition (IAEA, 2012a). The applicant requested to add a new content (i.e., JRR-4 HEU) to the package (see Section 1.1.2 of this SER). The addition of this content does not have any effect on the existing thermal evaluation of for the package. A summary of the staffs review is provided below.

4.1 Description of the Thermal Design The Model No. JCR-80Y-20T is a Type B(U) package designed to transport spent research reactor fuel. The addition of JRR-4 HEU fuel does not affect the thermal design of the package.

4.2 Thermal Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions The applicants previous evaluation bounds the new contents. Therefore, no additional review of the existing analysis was necessary.

4.3 Evaluation Findings

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

JCR-80Y-20T package, the staff concludes that the previous analyses of the thermal design of the JCR-80Y-20T package are bounding for the proposed contents. Therefore, the package meets the thermal requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a).

5.0 CONTAINMENT EVALUATION The purpose of the containment review is to verify that the package design satisfies the requirements for the evaluation of the containment boundary as required in the IAEA SSR-6, 2012 Edition (IAEA, 2012a). The staff confirmed that the containment system of the JCR-80Y-20T package is well described for revalidation.

6.0 SHIELDING EVALUATION The purpose of the shielding evaluation is to ensure the package meets the radiation level requirements within the IAEA Safety Standards (IAEA, 2012a and IAEA, 2012b) for protecting people and the environment. Table 5.1 below provides the applicable IAEA requirements used by the staff for evaluation of this package.

5 Table 6.1: Applicable IAEA, SSR-6, 2012 Edition, Shielding Requirements to Evaluate the Adequacy of a Package Design [Type B(U) Package].

Paragraph No. of the IAEA Brief Description of the Requirement SSR-6, 2012 Edition 526 The transport index (TI) shall not exceed 10.

The radiation level cannot exceed 0.1 milliSievert per hour 523 (mSv/hr) [10 millirems per hour (mrem/hr)] at 1 meter from the package.

527 (for non-exclusive use The maximum radiation level at the surface of the package packages) does not exceed 2 mSv/hr (200 mrem/hr).

652 (for Type B(U) The package shall meet the requirement in paragraph 648 of packages) SSR-6.

The package must not experience more than a 20% increase 648(b) (under NCT) in the maximum radiation level.

The package does not exceed 10 mSv/hr (1,000 mrem/hr) at 659(b)(1) 1 meter under accident conditions.

6.1 Shielding Design The applicant requested to add High Enrichment Instrumented Type fuel element from the JRR-4 Reactor for the Box-Type Basket previously approved by NRC in Revision 1 of Certificate No. 71-3035.

The staff reviewed the information related to the shielding components of the packaging, which consist mainly of a stainless-steel body and lid. The thickness of the package body is 31 centimeters (cm) in the radial direction and 30 cm in the bottom direction. The thickness of the lid is 37 cm. Also, there is shielding provided by the basket frame and the basket bottom plate made of stainless steel, and the structural materials of the fuel meat and the fuel cladding.

Six different fuel elements are allowed for transport within the JCR-80Y-20T. Five of the fuel elements were previously approved by the NRC in Revision 1 for this package. Table D.2 of the application includes the specifications of the fuel element such as burnup, enrichment, and cooling times used for the shielding analysis.

6.2 Shielding Evaluation The applicant calculated the gamma ray and neutron source terms using ORIGEN-JR code.

The ORIGEN-JR, is a computer code developed for calculating radiation sources and analyzing nuclide transmutations based on the ORIGEN code. The ORIGEN-JR code calculates the energy spectra of neutron sources from spontaneous fission and alpha and neutron (, n) reactions. The energy spectra calculated by the ORIGEN-JR code are the input to the shielding calculation code, DOT 3.5.

In Section D.2.1 of the application, the gamma and neutron source terms in Table D.3, and Table D.4 of the application, respectively, show that the JRR-4 HEU fuel is bounded by the JRR-4 low enrichment silicide fuel type within the Box Type Basket previously approved by NRC in Revision 1.

The shielding models are the same as the actual package configuration with the following exceptions:

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a. homogenization of the fuel region with the fuel meats, the fuel claddings, the structural materials of fuels,
b. neglect the fins on the surface of the package, and
c. ignore the neutron absorber.

The shielding model considered tolerances related to the plate dimensions in a conservative way that provides minimum shielding. Figure B.2 of the application depicts the main dimensional tolerances of the basket for Box Type Fuel. The applicant used the following source terms:

a. Gamma Source Term. The JRR-3 standard silicide type fuel has the maximum gamma source intensity per unit length contained in the basket for box type fuel.
b. Neutron Source Term. The JRR-4 low enrichment silicide type fuel has the maximum neutron source intensity per unit length contained in the basket for box type fuel.

The applicant evaluated the shielding capability of the package by calculating the external radiation level for the six kinds of fuel elements allowed for transport in the packaging. The applicant also evaluated the neutron source intensity used in the analysis by considering the effective multiplication factor (keff) obtained by using the three-dimensional multigroup Monte Carlo KENO-Va code.

The applicant evaluated the secondary gamma ray induced by neutrons at the same time as the neutron source terms. The applicant performed the gamma shielding calculation by using the point kernel code QAD-CGGP2R. This code is commonly used for shielding design of reactors, nuclear facilities, and transport casks of nuclear fuel and radioactive materials. The QAD-CGGP2R uses a different method for performing the flux-to-dose rate conversion than previously evaluated by the staff. The document JAERI-M 90-110, QAD-CGGP2 and G33-GP2: Revised Versions of QAD-CGGP and G33-GP (Codes with the Conversion Factors from Exposure to Ambient and Maximum Dose Equivalents), Y. Sakamoto, S. Tanaka, June 1990 (ADAMS Accession No. ML18270A163), Appendix A, Table A.2, includes a comparison of the flux-to-dose rate conversion factors for the ANS/ANSI-6.1.1 standard, Neutron And Gamma-Ray Fluence-To-Dose Factors, (ANS/ANSI-6.1.1 standard) and the QAD-CGGP2R code. The ANS/ANSI-6.1.1 standard conversion factors that is acceptable to the staff can be 40% higher than flux-to-dose rate conversion factors resulting from the method used by the applicant.

Although the applicants method is non-conservative, the staff found it acceptable for this application given the large margins (< 40%) to the radiation level limits in ANS/ANSI-6.1.1 standard.

The applicant performed the gamma streaming calculation using an analytical method in combination with the QAD-CGGP2R code. Section D.6.2 in Appendix 2 of the application includes an explanation of the analytical method used by the applicant. In this analytical method, since the neutron source intensity is very small (9.6x101 n/s/cm), the effect of neutron streaming is ignored. The applicant developed analytical models for the following three positions where the gamma streaming will have influence on the dose rate:

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a. drain hole, For the drain hole, the details of analysis are provided in Section 6.2.2.1 of the application. Figure D-42 of the application provides the modeling of the drain hole. The applicant developed the model to perform the analysis in accordance with the following steps:

(1) Calculate gamma flux at drain hole inlet point-1 by using QAD-CGGP2R code; (2) Calculate attenuation from plane source of point-1 to point-2; (3) Calculate gamma flux of scattering at point-2 and reaches point-3; and (4) Calculate attenuation of gamma flux by drain valve protection cover.

From the result above, the gamma streaming dose rate on the surface of the drain hole is less than 1 Sv/h (0.1 mrem), which does not represent a significant change in dose.

Similarly, the dose at 1 meter (m) from the surface, does not result in a significant change.

b. gap between the body and the lid, and For the gap between the body and the lid, Figure D.2.3 of the application includes the profile of the gap between the body and the lid. The applicant assumed that there are three routes as shown in Figure D.2.4 [i.e., Cases (A), (B) and (C)] as the pathway of leakage through the gap. In Case (A), there is no concern with the change as the shielding thickness is increased from 31 cm to 38.6 cm. In Case (B), there is no gap between the body and the lid, but the shielding thickness of the gasket groove on this route is 28.18 cm. The depth of gasket groove is less than 7 millimeters (mm). The gamma streaming dose rate of the route is calculated by using the QAD-CGGP2R code.

Figure D.2.5 of the application shows the analytical model. From the calculation, the gamma streaming dose rates of this route are determined to be as follows:

(1) at the surface of the package 1.1x101 microSievert per hour (Sv/hr)

(1.1 mrem).

(2) at 1 m from the surface of the package 1.3x100 Sv/hr (0.13 mrem).

c. vent hole.

Figure D.2.1 shows the profile of the vent hole. Figure D.2.7 of the application includes the model used to perform the analysis of the vent hole. The applicant followed these steps when performing the analysis:

(1) Calculate gamma flux at the vent hole inlet (point-P1) by using QAD-CGGP2R code; (2) Calculate the attenuation from the plane source of point-P1 to point-P2;

8 (3) Calculate gamma flux of scattering at point-P2 and reaches point-P3; and (4) Consider the attenuation of gamma flux by the vent valve protection cover.

The gamma streaming dose rate on the surface of the vent hole is determined to be 3Sv/h (0.3 mrem), which does not result in a significant change in dose. Similarly, the dose at 1 meter (m) from the surface, does not result in a significant change.

The staff examined the analytical method used by the applicant and found it acceptable because the applicant analyzed all potential positions where there could be gamma streaming.

The staff found that there is a very low dose rate on the surface and at 1 m from streaming at those positions.

The neutron shielding calculation is performed by using the two-dimensional discrete ordinates transport code DOT 3.5. The DOT 3.5 code uses the method of discrete ordinates to solve radiation transport problems in two spatial dimensions. While suitable for neutral particle transport analysis in fission reactor, CTR, and weapons studies, DOT 3.5 has been especially developed for large shielding problems involving the transport of neutrons and/or photons. The applicant performed dose rate calculations of each basket considering the fuel element with the maximum source terms. This method for evaluating the radiation level from the neutron source was accepted by NRC in Revision 1 to the JRC-80Y-20T package (ADAMS Accession No. ML18270A163).

The applicant notes that the structural and thermal analysis showed that the damage or deformation of the package by the drop test under the normal and accident conditions of transport are limited to local deformation of the fins. Even in the thermal test, no shielding material melts. Since the fins are neglected at the surface of the packaging in the shielding model, there is no reduction on the thickness of shield, even if there is deformation of the fins.

Therefore, the dose equivalent rate under the normal and accident conditions of transport is the same as that of routine conditions of transport.

Tables D.13 and D.14 include the radiation level results for each basket. The maximum dose equivalent rate at the top of the surface of package during normal conditions of transport is 218 Sv/hr (21.8 mrem/hr) and 58 Sv/hr (5.8 mrem/hr) at a point of 1 m from the top of the surface of the package. The dose satisfies the requirement of 200 mrem/hr or less and 10 mrem/hr or less, respectively, as specified in SSR-6 standard.

The maximum dose rate at a point of 1 m from the surface of the package under accident conditions of transport is 0.058 mSv/hr (5.8 mrem/hr), which also satisfies 10.0 mSv/hr (1,000 mrem/hr) or less specified in SSR-6 standard. The gamma streaming dose equivalent rate is 11 Sv/hr (1.1 mrem/hr) at the gap between the body and the lid, which shows the maximum value.

Based on information provided in the application and that the JRR-4 HEU fuel has a lower burnup and a larger cooling time in comparison to the other fuels authorized for transport in the Model No. JCR-80Y-20T, the staff concluded that the proposed content is reasonably bounded by the other fuels. In terms of the dose rates associated with the JRR-4 HEU fuel, the staff found that the calculated radiation levels are under regulatory limits in paragraph No. 527 of the SSR-6, 2012 Edition (IAEA, 2012a).

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6.3 Evaluation Findings

Based on review of the statements and representations in the JCR-80Y-20T package application and as discussed in the paragraphs above, the staff has reasonable assurance that the JCR-80Y-20T package meets the shielding safety requirements in Paragraphs 526, 527, 648(b), and 659(b)(1) in IAEA SSR-6, 2012 Edition. The staff recommends revalidation of Japanese Certificate of Approval No. J/61/B(U)F-96, Revision 3, for the Model No. JCR-80Y-20T package.

7.0 CRITICALITY SAFETY EVALUATION The staff reviewed the application for revalidation of the JRC-80Y-2OT Type B(U) package to ensure that the package will continue to remain subcritical under the requirements of IAEA regulation SSR-6, 2012 Edition (IAEA, 2012a). The staff specifically focused the review of the addition of the JRR-4 HEU fuel elements as an allowable fuel type in the package against the respective paragraphs of SSR-6 listed in the table below.

Table 7.1: Applicable IAEA, SSR-6, 2012 Edition, Criticality Requirements to Evaluate the Adequacy of a Package Design [Type B(U) Package].

Paragraph No. of the Brief Description of the Requirement SSR-6, 2012 Edition 501(c) Ensure criticality safety features are within the limits of the design 525 Determination of CSI for shipments 673 Requirements for packages containing fissile material 680 Assessment of package in isolation 682 Package subcritical under paras 680 and 681 685 Arrays of packages 686 Determination of CSI for packages 7.1 Description of Criticality Design The applicant requested to add JRR-4 HEU fuel elements as an allowable fuel type in standard Box type basket previously approved by the NRC in Revision 1 of the Japanese Certificate No.

71-3035. This new fuel type consists of fuel elements similar to the JRR-3 and JRR-4 (low enrichment fuels already approved for this package in the standard Box type basket.

The design of the JRC-80Y-20T package remains unchanged from the previous package and basket design approved by the NRC in Revision 1 of CoC No. 3035. The applicant continues to use the boral neutron poison that has been approved for up to 75% credit of the entrained boron.

7.2 Spent Nuclear Fuel Contents The JRR-4 HEU fuel are a plate assembly design similar to the fuels already allowed in the Box type basket by Revision 1 of CoC No. 3035, but the Uranium-235 (235U) enrichment is increased to 93% and the mass of 235U is reduced to 168 grams (g) per fuel element for the new JRR-4 HEU fuel. The JRR-4 HEU fuel consists of aluminide and silicide plate assemblies. Prior to packaging the fuel in the JRC-80Y-2OT package, the JRR-4 HEU fuel has the portions of the assembly that do not contain fuel removed. The fuel is 84.0 cm long with the top 21.5 cm and

10 the bottom 2.5 cm of length retained as structural components to maintain the geometry of the fuel assembly.

7.3 General Consideration for Criticality Safety The previous revalidation of CoC No. 3035 (Revision 1) allowed for up to 40 assemblies of JRR-3 and JRR-4 (low enriched) fuels to be transported in the package. The applicant used the same methodology for modeling the new contents in the JRC-80Y-2OT package as was used previously, and staff determined that this approach was applicable to the new JRR-4 HEU fuel type because the applicant applied appropriate credit for the borated plates, the fuel assemblies were explicitly modeled, and the basket design was unchanged for the new fuel contents.

7.3.1 Evaluation of Single package The applicant assumed a single package filled with optimum moderation and with full water reflection for their analysis. Optimum moderation in this case, as well in the arrays, was full density water in the cask cavity. Since this condition is the same as the package array evaluation, the applicant indicated that this configuration is bounded by the array evaluation, which is much more conservative. The staff finds this assessment to be acceptable, since the package body, lid, and base are stainless steel, the keff of the of the single package with the containment system reflected is bounded by the infinite array of packages.

7.3.2 Evaluation Array of packages The applicant analyzed an infinite array of packages with optimal internal moderation (i.e., full density water) for the Model No. JRC-80Y-2OT package loaded with JRR-4 HEU fuel assemblies. The applicant determined for the JRR-4 HEU fuel that the maximum keff , including uncertainties, equaled 0.699, much less than the bounding JRR-3 fuel with a keff , including uncertainties, of 0.873.

7.3.3 Criticality Safety Index (CSI)

The applicant determined that the CSI for the JRC-80Y-2OT package remained at zero with the new proposed contents of JRR-4 HEU fuel assemblies because the applicant demonstrated that an infinite array of packages under normal and hypothetical accident conditions of transport remain subcritical. The staff finds that the applicant correctly calculated the CSI and, therefore, it is acceptable.

7.4 Demonstration of Maximum Reactivity The staff evaluated the proposed JRR-4 HEU fuel type as an allowable content in the JRC-80Y-2OT package in the standard Box type basket and compared it to the previously approved JRR-3 silicide fuel analysis. The applicant performed calculations using the SCALE system of codes using the KENO-Va Monte Carlo calculation code and the ENDF/B-V 238-group data library for the new HEU JRR-4 fuel using similar modeling techniques to evaluate normal conditions of transport and hypothetical accident conditions using optimum moderation, water intrusion, and fuel assembly tolerances. In all instances, the applicants calculated keff were substantially below the bounding multiplication factor of the JRR-3 fuel analysis. The keff of the package with the new content is 0.699 ( keff + 3 = 0.699) for the HEU JRR-4 fuel versus keff + 3 = 0.873 for the JRR-3 fuel. The staff performed confirmatory calculations based on the new JRR-4 HEU fuel contents using SCALE 6.1 computer code system, with the KENO VI three-dimensional

11 Monte Carlo code, and the continuous energy ENDF/B-VII cross-section library using assumptions similar to the applicants. The staffs calculations confirmed the applicants results for the JRR-4 HEU fuel type in the JRC-80Y-2OT package. The staff determined that the increase in enrichment was offset by the large reduction in uranium mass of the JRR-4 HEU fuel, resulting in a lower keff. The staff finds that the maximum reactivity is bounded by the analysis of the JRR-3 fuel type.

7.5 Evaluation Findings

The staff found that the proposed addition of HEU JRR-4 fuel in the JRC-80Y-2OT package will remain subcritical for all routine, normal, and accident conditions of transport. The staff based its finding on its verification of adequate system modeling performed by the applicant and that the JRR-4 HEU fuel is bounded by previous analysis (i.e., CoC No. 3035, Revision 1). The acceptance standard of a maximum keff of 0.95 was maintained for all analyzed scenarios and meets the requirement that the package maintain subcriticality under all conditions of routine, normal and accident conditions as required by the IAEA SSR-6, 2012 Edition, paragraphs 637(a) and 682.

8.0 CONDITIONS The staff did not recommend additional conditions for revalidation of Japanese Certificate of Approval No. J/61/B(U)F-96, Revision 3, for the Model No. JCR-80Y-20T package.

9.0 REFERENCES

(IAEA, 2012a) 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.

(IAEA, 2012b) International Atomic Energy Agency, IAEA SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2012 Edition, https://www-pub.iaea.org/MTCD/publications/PDF/Pub1586web-99435183.pdf.

(NRC, 1997) Leeds, Eric, U.S. Nuclear Regulatory Commission (NRC,) letter to Boyle, Richard W., U.S. Department of Transportation (DOT),

July 29, 1997, ADAMS Package Accession No. ML023080351.

(DOT, 2018) Boyle, Richard W., U. S. Department of Transportation (DOT),

letter to Mr. Michael Layton, U.S. Nuclear Regulatory Commission (NRC), August 8, 2018, ADAMS Package Accession No. ML18270A151.

(DOT, 2019) Boyle, Richard W., U. S. Department of Transportation (DOT),

letter to N. García Santos, U.S. Nuclear Regulatory Commission (NRC), February 19, 2019, ADAMS Package Accession No. ML19059A420.

12 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. JCR-80Y-20T package meets the requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a).

Issued with letter to R. Boyle, U. S. Department of Transportation, on 8/1/19.

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