ML19269C805

Certificate of Approval No. F/357/B(U)F-96, for the Model No. TN-MTR Package - Revalidation Recommendation

ML19269C805
Person / Time
Site:	07103052
Issue date:	10/11/2019
From:	John Mckirgan Division of Spent Fuel Management
To:	Boyle R US Dept of Transportation, Radioactive Materials Branch
BHWhite NMSS/DSFM 415.6577
References
EPID L-2019-LLA-0044
Download: ML19269C805 (17)

Text

October 11, 2019 Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 1200 New Jersey Avenue SE Washington, D.C. 20590

SUBJECT:

CERTIFICATE OF APPROVAL NO. F/357 /B(U)F-96, FOR THE MODEL NO.

TN-MTR PACKAGE - REVALIDATION RECOMMENDATION

Dear Mr. Boyle:

This is in response to your letter dated March 5, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19115A127), as supplemented on June 12, 2019 (ADAMS Accession No. ML19169A036), and September 19, 2019 (ADAMS Accession No. ML19266A521), the United States Department of Transportation requested that the U.S.

Nuclear Regulatory Commission (NRC) staff perform a review of the French Certificate of Approval No. F/357/B(U)-96 Revision Eah, for the Model No. TN-MTR transport package.

Specifically, you requested that the NRC make a recommendation concerning the revalidation of the package for import and export use. Specifically, you requested that the NRC only review the content in Appendix 9 to the French certificate for shipment of the Gisete 4, Gisete 5, or Gisete 8 radioisotopic thermal generators.

Based upon our review, the statements and representations contained in the application, in the Package Design Safety Report No. DOS-18-011415-000, Version 2.0, as supplemented, and for the reasons stated in the enclosed safety evaluation report, we recommend revalidation of French Certificate of Approval No. F/357/B(U)-96, Revision Eah transport package.

R. Boyle If you have any questions regarding this matter, please contact me or Bernard White of my staff at (301) 415-6877.

Sincerely,

/RA/

John McKirgan, Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3052 EPID L-2019-LLA-0044

Enclosure:

Safety Evaluation Report

ML19269C805 OFFICE: DSFM DSFM DSFM DSFM DSFM SFigueroa YKim JBorowsky JWise NAME: BWhite Via email Via email Via email Via email DATE: 9/19/19 9/23/19 9/20/19 9/24/19 9/19/19 OFFICE: DSFM DSFM DSFM DSFM DSFM VWilson TTate MRahimi YDiaz-Sanabria NAME: JMcKirgan Via email Via email Via email Via email DATE: 9/19/19 9/26/19 9/26/19 9/26/19 10/11/19 SAFETY EVALUATION REPORT Docket No. 71-3052 Model No. TN-MTR French Certificate of Approval No. F/357 /B(U)-96 Revision Eah

SUMMARY

By application dated March 5, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19115A127), as supplemented on June 12, 2019 (ADAMS Accession No. ML19169A036), and September 19, 2019 (ADAMS Accession No. ML19266A521), the United States Department of Transportation (DOT) requested that the U.S.

Nuclear Regulatory Commission (NRC) staff perform a review of the French Certificate of Approval F/357/B(U)-96 Revision Eah, for the Model No. TN-MTR transport package.

Specifically, you requested that the NRC make a recommendation concerning the revalidation of the package for import and export use and that the NRC only review the content in Appendix 9 to the French certificate for shipment of the Gisete 4, Gisete 5, or Gisete 8 radioisotopic thermal generators (RTGs).

In support of this request the DOT provided the following documents with its letter dated September 26, 2018:

1. French Certificate of Approval No. F/357 /B(U)-96, Revision Eah, dated December 26, 2018.

2. Package Design Safety Report No. DOS-18-011415-000, Version 2.0, as supplemented.

The NRC previously reviewed and recommended revalidation of this package to the DOT on March 19, 2004 (ADAMS Accession No. ML040820100) and July 10, 2019 (ADAMS Accession No. ML19 ML19192A055). Based on the statements and representations in the information provided by DOT and TN Americas LLC, the NRC staff recommends that French Certificate of Approval No. F/357 /B(U)-96, Revision Eah, dated December 26, 2018, be revalidated for the contents listed below (see Section 1.2, Contents).

1.0 GENERAL INFORMATION 1.1 Package Description The TN-MTR package is designed to transport various research reactor spent fuel, including material testing reactor (MTR) and a Gisete 4, Gisete 5, or Gisete 8 RTG, which is placed in a specific internal fitting. The packaging consists of three main components: the package body, the impact absorbing cover, and the lid. The package body is constructed of stainless steel and incorporates lead gamma shield and thermal insulation for protection against fires. Fins are welded to the external shell of the packaging. The top part of the body includes a flange that has tapped holes for the lid and the lid screws. Two trunnions are screwed onto the top flange.

Enclosure

Tie-down lugs are welded to the outer shell. The impact absorbing cover, or impact limiter, consists of wood blocks encased in a steel skin, and is attached to the package body by screws.

The package is closed by a stainless steel lid that also incorporates lead for radiation shielding and is equipped with two ports for draining and drying. The lid is fixed to the package body by 36 screws and has double elastomeric O-ring seals for containment and to facilitate leak-testing.

A removable internal fitting within the package cavity accommodates a single strontium titanate (SrTiO3) RTG. There are 3 types of internal fittings, each dedicated to one of the three Gisete RTGs. The internal fittings for the generators consist of an axial wedge and a radial wedge that can be fitted with a bottom.

The Gisete RTGs consist of a metal capsule locking the radioactive material; which forms the source block; a tungsten alloy or depleted uranium primary shield in which the source block is inserted; a ceramic thermal insulator; a secondary shield constituting the bodys outer plate; and a closure system that can contain an electricity generation block (the block is absent from Gisete 8). In addition, these different components can be equipped with attachment systems (screws, tie-rods, etc.) or gaskets.

The package has the following approximate dimensions and weights:

Overall package diameter, with impact limiter 2080 mm Overall package height, with impact limiter 2008 mm Package cavity diameter 960 mm Package cavity height 1080 mm Packaging mass, without basket and contents 20,600 kg Maximum package mass, with contents 23,400 kg 1.2 Contents The contents of the package consiste of a Gisete 4, Gisete 5, or Gisete 8 RTG, which may contain a maximum of 897 TBq of Strontium-90 (Sr-90), producing less than or equal to 160 W of decay.

2.0 STRUCTURAL EVALUATION The objective of the structural evaluation is to verify that the structural performance of the package has been evaluated to meet the regulatory requirements of the International Atomic Energy Agency (IAEA) Safety Standards Series, No. SSR-6; Regulations for the Safe Transport of Radioactive Material, 2012 Edition.

2.1 General Consideration The TN-MTR packaging is designed to transport MTR fuel from research reactors. It is also used for the transport of sealed sources with a special shape placed in the packaging cavity by a special internal fitting.

The applicant submitted a revision to revalidate the French certificate for contents consisting of an RTG with types: Gisete 4, Gisete 5 or Gisete 8, placed in a special internal fitting, as described in the French Certificate of Approval No. F/357/B(U)F-96 (Eah), Appendix 16.

Therefore, this safety evaluation report (SER), Section 2.0 evaluates structural design and

analysis of the Gisete 4, Gisete 5 and Gisete 8 RTGs, as described in Appendix 16 of the French certificate.

2.2 Packaging Masses and Centers of Gravity In Chapter 0A-Appendix 14: Description of the Gisete Content and Its Specific Internal Fitting of the SAR (Document No. DOS-18-011415-041, Ver. 1.0), the applicant stated that the maximum masses of the Gisete 4, Gisete 5 and Gisete 8 RTGs are 2,270 kg, 1,670 kg and 212 kg, respectively. In addition, maximum masses of the internal fittings for the Gisete 4, Gisete 5 and Gisete 8 RTGs are 255 kg, 205 kg and 225 kg for the axial wedge and 155 kg, 295 g, and 260 kg for the radial wedge, respectively. Since the maximum potential mass of the Gisete RTG with the internal fitting is less than the maximum allowable mass for the packing of 2,800 kg, the mass increase by the Gisete RTG with its internal fitting is acceptable.

Additionally, the staff evaluated that the center of the gravity (CG) of the Gisete 4 RTG, which has the largest dimension among three RTGs, is almost identical to the CG of the packaging with the maximum allowable mass of 2,800 kg.

Structurally, the TN-MTR package has not significantly changed in overall dimensions, CG, weight, and construction from the previous SAR version, which the staff previously reviewed and provided a recommendation to DOT to revalidate the package, except as noted here; therefore, the TN-MTR package will experience the same g-loads as observed in the previously approved SAR version. Only those exceptions (internal fittings of the Gisete 4, Gisete 5 and Gisete 8 RTGs) that affect the structural performance of the package have been reviewed and evaluated.

2.3 General Description of the Radioisotopic Thermal Generator Internal Fittings The RTG is wedged in the TN-MTR cavity axially and radially by means of an internal fitting.

There are three types of internal fittings: Gisete 4 internal fitting to wedge the Gisete 4 RTG, Gisete 5 internal fitting to wedge the Gisete 5 RTG, and Gisete 8 internal fitting to wedge the Gisete 8 RTG. The internal fittings have 2 parts: (i) a radial wedge that can be fitted with a bottom, whose function is to radially wedge the Gisete RTG in the TN-MTR cavity, and (ii) an axial wedge, whose purpose is to axially wedge the Gisete RTG in the TN-MTR cavity and reduce the axial gap between the internal fitting and the lid. The drawing of the internal fittings is shown in Appendix 0A-14-1 of the Document No.

DOS-18-011415-042, Version 1 of the SAR and details of the description is provided in Chapter 0A-Appendix 14, Description of the Gisete Content and Its Specific Internal Fitting of the SAR.

The structural parts of the RTG internal fittings are made from Type A aluminum. The mechanical characteristics of this material are given in Table 0A-14.2 of the SAR and summarized below:

• Yield strength (y) 115 MPa @20 °C and 100 MPa @100 °C,

• Ultimate tensile strength (u) 270 MPa @20 °C,

• Modulus of elasticity (E) = 71,000 MPa @25 °C and = 68,000 MPa @100 °C,

• Poisson's ratio (v) = 0.33, and

• Thermal expansion = 23.4.10-6 K-1 @100 °C.

2.4 Materials Evaluation The staffs materials review focused on the chemical compatibly and mechanical properties of the new RTG contents and internal fittings. The service conditions for the other, unchanged, packaging components (e.g., seals, thermal insulation, metal containment) are bounded by the conditions considered in the staffs prior reviews of the TN-MTR package, and thus the materials performance of those components is not evaluated.

2.4.1 Chemical Compatibility and Radiolysis As described in Chapter 0A-14 of the SAR, the internal fittings that position the RTGs within the TN-MTR packaging cavity are constructed of an aluminum alloy. These fittings are in contact with the stainless steel internal containment liner of the packaging and the exterior housings of the RTGs. Depending on the RTG model, the external surfaces of their housings may be lead, cast iron, or other metallic materials. As stated in SAR Chapter 6A, the RTGs must be loaded in a dry condition. The containment filler gas can be air, helium, or any inert gas.

The staff reviewed the materials and service environments associated with the new RTG contents to determine if the applicant adequately accounted for the potential for corrosion or other adverse reactions. The staff considers the packages containment seals and the loading of the contents in only a dry condition to be effective means to ensure that there is not sufficient water present to support corrosion within the packaging cavity. The RTG housings are not relied on to support a safety function, and the staff finds that, with the exclusion of water from the cavity, the contacts between the aluminum internal fittings and metallic RTG housings do not introduce adverse corrosive reactions that could credibly compromise the structural performance of the fittings. The staff also notes that, in accordance with the maintenance program outlined in Table 7A.1 of the SAR (Document No. DOS-18-011415-036, Rev 1.0), the overall condition of the packaging internals is visually assessed during each usage.

The staff also evaluated for the potential of radiolysis of water present in humid air within the packaging cavity. Even assuming complete radiolysis of the available water in the air, the staff finds that the resulting hydrogen generation could not reach levels to support flammability.

Based on the evaluations above, the staff finds that the applicant adequately accounted for chemical compatibility and other adverse reactions, and the package meets the requirements in Paragraphs 507, 614, and 644 of SSR-6.

2.4.2 Mechanical Properties As stated in Section 5 of Chapter 1A-14 of the SAR (Document No. DOS-18-011415-043 Version 1.0), the internal fittings that position the RTGs within the TN-MTR packaging cavity are necessary to ensure that shifting of the RTGs in an accident will not damage the packaging.

The fittings are not relied on for positioning the RTGs, as the shielding calculations assume the Sr-90 source is conservatively placed adjacent to the packages internal cavity liner. The fittings are constructed of aluminum alloy grade 5083, conforming to either European Standard EN 482-2 or ASTM International Standard B209M. As stated above in Section 2.3, SAR Table 0A-14.2 provides the mechanical properties of the aluminum, including a minimum yield strength of 100 MPa (14.5 ksi). The applicant cites ASTM and American Society of Mechanical Engineers (ASME) standards for the sources of the materials properties.

The staff reviewed the mechanical properties of the aluminum in the applicants structural analyses to verify that the properties adequately account for the service environment. As shown in Table 2A.1 of the SAR (Document No. DOS-18-011415-028 Vers. 2.0), the applicant conservatively assumed that internal fittings may reach 100°C (212°F) in normal conditions of transport. The staff notes that aluminum alloys may lose strength with prolonged exposure to elevated temperature. Thus, the staff evaluated whether the strength of the internal fittings could fall below the values assumed in the structural analyses. The staff notes that laboratory data on the strength of annealed 5083 shows that, after exposure for 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 100°C (212°F), the yield strength is approximately 145 MPa (21 ksi) [ASM, 1998]. As a result, the staff finds that applicants use of a 100 MPa (14.5 ksi) yield strength in the structural analyses conservatively accounts for the elevated temperature exposure. The staff also reviewed the tensile strength, elastic modulus, Poissons ratio, and thermal expansion coefficient in SAR Table 0A-14.2 and verified that the properties used in the applicants structural analyses are consistent with those in the technical literature over the temperature range in which the package must be demonstrated to perform.

The staff also reviewed the radiation levels from the RTGs with respect to potential effects on material properties. The staff notes that the radiation from the RTGs is significantly below levels evaluated by the staff in prior reviews of the packaging for other content types (e.g., research reactor fuels) and are significantly below levels capable of degrading aluminum properties. As a result, the staff finds that radiation will not detrimentally affect the properties of the packaging materials.

Based on the above evaluations of temperature and radiation effects, the staff finds that the applicant adequately accounted for the effects of service conditions on the mechanical behavior of the packaging materials. The staff concludes that the package meets the requirements in Paragraphs 614, 639, and 640 of SSR-6.

The staff reviewed the adequacy of the materials used in the TN-MTR package for the transport of the RTG contents and finds that the package design property accounts for the chemical compatibility of package components, potential adverse reactions, and mechanical performance of packaging materials under the loads and environments required for evaluation in the IAEA SSR-6 (2012) regulations.

Reference ASM, Aluminum and Aluminum Alloys: Properties of Wrought Aluminum Alloys, Metals Handbook, Desk Edition, 2nd Edition, ASM International, Materials Park, Ohio, 1998.

2.5 Structural Analysis of the Radioisotopic Thermal Generator Internal Fittings for Normal Conditions of Transport The applicant analyzed the three Gisete RTG internal fittings for the normal conditions of transport using a linear elastic formula to calculate compressive stress of the Gisete RTG fittings. The 0.3 m lateral drop was considered and an input of the transverse acceleration of 22.5g was obtained from the previous model tests. A vertical drop was not considered because it is bounded by the accident conditions of transport.

In Section 4.3 of Chapter 1A-Appendix 14 of the SAR, the results of the analysis show that the calculated compressive stress of the internal fittings for the Gisete 4, Gisete 5 and Gisete 8 RTGs are 29 MPa, 31 MPa and 16 MPa, respectively. Since these compressive stresses are

smaller than the yield stress of 100 MPa, the design of the Gisete RTG internal fittings under the normal conditions of transport is acceptable.

2.6 Structural Analysis of the Radioisotopic Thermal Generator Internal Fittings for Accident Conditions of Transport The applicant analyzed the three Gisete RTG internal fittings for the accident conditions of transport using a linear elastic formula to calculate compressive stress of the RTG fittings. The 9.0 m vertical drop was considered and an input of the vertical acceleration of 149.0g was obtained from the previous model tests.

In Section 5.3 of Chapter 1A-Appendix 14 of the SAR, the results of the accident conditions of transport analysis show that the calculated compressive stress of the internal fittings for the Gisete 4, Gisete 5 and Gisete 8 RTGs are 68 MPa, 55 MPa and 36 MPa, respectively. Since these compressive stresses are smaller than the yield stress of 100 MPa, the design of the Gisete RTG internal fittings under the accident conditions of transport is acceptable.

2.7 Thermal Expansion Analysis of the Radioisotopic Thermal Generator Internal Fittings The applicant also analyzed the thermal expansion of the three Gisete RTG internal fittings due to the temperature increase to make sure that the expansions do not close the gap between the internal fittings and the cask cavity, thereby, potentially creating additional stresses.

The applicant used a linear elastic formula to calculate thermal expansions in the axial and radial directions. The initial and final temperatures of 20 °C and of 100 °C, respectively, were considered and the coefficient of thermal expansion of 23.4x106 K1 was used for the Type A aluminum for the thermal expansion analysis.

The applicant provided the results of the analysis in Section 6 of Chapter 1A-Appendix 14 of the SAR. The results show that the calculated axial expansion of 2.0 mm of the internal fittings, which is smaller than the minimum gap of 4.0 mm between the internal fittings and the lid. The results also show that with the calculated radial expansion of 0.9 mm of the internal fittings, the radial wedge maximum size is smaller than the minimum radius of the inner cavity. Since these thermal expansions are smaller than the minimum design gaps, the design of the Gisete RTG internal fittings under the temperature increase is acceptable.

2.7 Evaluation Findings

Based on the review of the statements and representations contained in the application, the staff concludes that the structural designs for the Gisete Types 4, 5 and 8 RTG internal fittings have been adequately described and evaluated and that the TN-MTR package has adequate structural integrity to meet the requirements of the IAEA SSR-6, 2012 Edition.

3.0 THERMAL EVALUATION 3.1 Description of Thermal Design The TN-MTR package is designed to transport MTR fuel elements and other radioactive material. The objective of the current review was to recommend revalidation of the TN-MTR package that allows transport of a Gisete 4, Gisete 5, or Gisete 8 RTG as content. Drawing PLA-15-00166-000 Rev. 0 and drawing PLA-15-00166811-001 Rev. 0 were the same from the

previous TN-MTR revalidation application with MTR fuel elements as content, indicating that the packaging did not significantly change. Discussion of the Gisete RTG internal fittings was provided in Chapter 0A-Appendix 14 and the internal fitting drawings for the Gisete 4, Gisete 5, and Gisete 8 RTG content were provided in drawings PLA-17-00204480-015-1.0, PLA 00204480-016-1.0, and PLA-17-00204480-017-1.0, respectively. Chapter 1A and Chapter 1A-Appendix 14 indicated that these internal fittings secure the Gisete content within the TN-MTR package and provide for sufficient wedging of the RTG during normal transport conditions and accident transport conditions.

Chapter 3A (page 19/35 of Document No. DOS-18-011415-029 Vers. 2.0) indicated that the Gisete content is loaded at atmospheric pressure (1.04 bar). An internal cavity temperature of 482 K at normal transport conditions (Chapter 3A - Appendix 15) would raise this pressure to 1.68 bar (Chapter 3A - Appendix 15, page 7/8). Likewise, with a cavity temperature of 522 K (Chapter 3A - Appendix 15, page 4/8) at the thermal hypothetical accident transport condition, the cavity pressure would rise to 1.82 bar (Chapter 3A - Appendix 15, page 8/8). These pressures are below the maximum design pressure of 7 bar reported on page 24/35 of Chapter 3A (and Chapter 1 and Chapter 1-10).

3.2 Thermal Evaluation under Normal Conditions of Transport According to the current revalidations Appendix 16 of French Certificate of Approval No.

F/357/B(U)F-96 (revision Eah), the decay heat of the Sr-90 content in a Gisete isotopic generator is no more than 160 W, which is less than the 260 W decay heat described in the above-mentioned revalidation with MTR fuel elements. Thus, temperatures and pressures would be reduced compared to those values reported for the MTR fuel element content. This is indicated in Chapter 2-Appendix 1 (Document No. DOS-18-011415-025 Rev. 1.0), which presented the boundary conditions and results for a three-dimensional model thermal analysis (using IDEAS 6.1 M1 / TMG Thermal Analysis Module/Thermal flow 6.0.1181) of a TN-MTR package content with 5500 W decay heat. Results indicated the ethylene propylene diene terpolymer (EPDM) seal temperature is less than the 160 °C allowable temperature, resin temperature is less than its allowable temperature, and lead temperature is less than the 327 °C allowable temperature. In addition, page 4/23 of Chapter 2 (Document No.

DOS-18-011415-024 Rev. 1.0) stated that changing from F resin to Vyal B resin had no influence on packaging temperatures under normal or accident transport conditions. Finally, the document Test Program: Acceptance and Maintenance (Document No. DOS-18-011415-036, Rev. 1.0) noted on page 5/12 that a thermal acceptance test, based on 8 kW being distributed within the package, is performed to verify the heat dissipation characteristics of the package.

3.3 General Considerations Document DS LME 50291001-01 Revision C (page 19/24) stated that Chapter 04-7 demonstrated the packages maximum accessible surface temperature is not greater than 50 °C, and thus, the package is shipped as non-exclusive use. This is further clarified in the response to the September 19, 2019, thermal response to NRCs request for additional information which indicated a 43 °C surface temperature. In addition, Document DS LME 50291001-01 Revision C (page 16/24) stated that Chapter 04-8 demonstrated that package components had allowable temperatures at -40 °C.

As noted in the French approval certificate, contained transport methods (e.g., covered vehicle, transport container, canopies) are not permitted unless the thermal power is such that critical

components of the package, such as the resin, elastomeric seals, and internal fittings, do not exceed their allowable temperatures.

It is noted there are some uncertainties associated with the equation and analyses in Results of the Thermal Test Report under Transport Conditions (Document No. DOS-18-011415-026),

including pages 5/18 and 6/18. However, the bounding nature of the modeled decay heat (1200 W and 5500 W versus the requested revalidations 160 W) indicates that component temperatures would be below their allowable values for content with a 160 W decay heat.

3.4 Thermal Evaluation under Hypothetical Accident Conditions Chapter 2 - Appendix 2 (Document No. DOS-18-011415-027 Rev. 1.0) presented the boundary conditions and results of hypothetical accident thermal condition analyses. Results indicated that with 5500 W decay heat content, the containment boundary seals would be 201.1 °C and lead would be 225.8 °C, both temperatures being less than the applicants stated allowable of 220 °C (short term seal allowable) and 327 °C (lead melting point). In addition, Chapter 2A-10 of the application noted that the thermal analysis for the CESOX content and its internal temperatures, which were based on a decay heat of 1200 W, bounds the Gisete content. The CESOX content thermal analysis in Chapter 2A - Appendix 10 (Document No. DOS 00173678-216, Rev. 2) showed that the maximum temperature of the lead was 173 °C after the thermal hypothetical accident transport condition. The above-mentioned temperatures, including the seals, would be reduced if the actual decay heat of 160 W was transported and, thus, there is reasonable assurance that the seals would maintain their integrity under accident conditions. Additional discussion of a seals performance relative to its allowable temperature is presented in the SERs Structural Evaluation Chapter 2, above.

Based on the relevant portions of the French certificate and the representations in the application, the staff evaluated the revalidation request for the TN-MTR package with Gisete RTG content and concludes that thermal-related provisions of the IAEA Specific Safety Requirements, No. SSR-6, 2012 edition will continue to be met.

4.0 CONTAINMENT EVALUATION The TN-MTR package is designed to transport MTR fuel elements and other radioactive material. The objective of the current review was to recommend revalidation for the TN-MTR package that would allow transport of a Gisete 4, Gisete 5, or Gisete 8 RTG as authorized content. Drawing Nos. PLA-15-00166-000 Rev. 0 and PLA-15-00166811-001 Rev. 0 were the same from the previous TN-MTR revalidation application with MTR fuel elements as content, indicating that the package did not significantly change. For example, the torques associated with the lid-to-body screws and orifice closure plate screws remained at 660 Nm and 40 Nm, respectively (Chapter 6A, Table 6A.1).

4.1 General Considerations The discussion of the containment boundary (comprised of the bottom of containment body, shell, flange, lid, inner seal of lid, orifice A plug cover and inner seal, and orifice B plug cover and inner seal, per Section 7 of Chapter 0 Description of the TN-MTR Packaging) was presented in the previously mentioned SER associated with the MTR fuel element as content (ML19192A055). Discussion of the Gisete RTG internal fittings was provided in Chapter 0A-Appendix 14 and the internal fitting drawings for the Gisete 4, Gisete 5, and Gisete 8 content were provided in Drawing Nos. PLA-17-00204480-015-1.0, PLA-17-00204480-016-1.0, and

PLA-17-00204480-017-1.0, respectively. According to Chapter 1A and Chapter 1A-Appendix 14, these internal fittings secure the Gisete RTG within the TN-MTR package and provide for sufficient wedging of the RTGs during normal transport conditions and accident transport conditions, which would result in the Sr-90 titanate SrTiO3 source, with an activity of 897 TBq, maintaining its sintered pellet form within the RTG.

Nonetheless, the release calculations provided in Chapter 3A assumed the content as an aerosol within the TN-MTR cavity even though no mechanism was presented which would result in a breach of the RTG and a changing of form of the SrTiO3 source from pellet to aerosol. For normal conditions, the release calculations were based on an aerosol concentration of 1E-3 g/m3; at accident conditions the calculations were based on 9 g/m3 aerosol concentration within the first 30 minutes of an accident and 0.1 g/m3 thereafter. Under these conditions and assuming a leakage rate of 4.7E-4 Pa m3/sec standard leak rate (SLR) (per page 16/35 of Chapter 3A), it was reported that the total release at normal conditions and accident conditions would be 2.69E-7 A2/hour and 5.99E-3 A2/week, respectively. These releases are below the 1E-6 A2/hour and A2/week regulatory limits at normal conditions and accident conditions, respectively.

4.2 Leakage Rate Testing According to Chapter 7A (page 4/12), packaging acceptance tests include fabrication leakage rate tests of the lid inner seal, orifice cover inner seals, and containment welds; the acceptance criterion is a total leakage rate of 3.5E-5 Pa m3/sec SLR, with the weld acceptance criterion being 1.1E-7 Pa m3/sec SLR. In addition, Chapter 7A lists the maintenance leakage rate tests associated with each shipment (i.e., cycle), after 15 shipments (or 3 years, whichever is more severe and occurs first), and 60 shipments (which also includes the leakage rate tests associated with the 15-shipment maintenance program). For each shipment, Table 7A.1 indicated that the lid and orifice cover inner seals are replaced (if necessary) and undergo a (pre-shipment) leakage rate test; Chapter 6A (page 30/34) stated the acceptance criterion of the sum of the leakage rate tests is 4.7E-4 Pa m3/sec SLR.

For the 15-shipment maintenance program, Table 7A.2 stated that the lid and orifice cover seals are replaced, and leak tested with a total leakage rate acceptance criterion of 3.5E-5 Pa m3/sec SLR. For the 60-shipment maintenance program, Table 7A.3 stated that the welds are leak tested with an acceptance criterion of 1.1E-7 Pa m3/sec SLR. According to Table 6A.3 of Chapter 6A, the leakage rate tests comply with International Organization for Standardization (ISO) Standard No.12807, Safe Transport of Radioactive MaterialsLeakage Testing on Packages and are conducted by qualified personnel.

It is noted that page 4/12 of Chapter 7A indicated leakage rate acceptance criteria of 3.5E5 Pa m3/sec SLR and 1.1E7 Pa m3/sec SLR, rather than 3.5E-5 Pa m3/sec SLR and 1.1E-7 Pa m3/sec SLR.

Based on the relevant portions of the French certificate and the representations in the application, the staff evaluated the revalidation request for the MTR package with Gisete content and concludes that containment-related provisions of the IAEA Specific Safety Requirements, No. SSR-6, 2012 edition will continue to be met.

5.0 SHIELDING EVALUATION The TN-MTR is a Type B(U) package and is designed for exclusive use. The staff reviewed the application to ensure that the shielding design of the package meets the radiation level requirements of the IAEA safety standards SSR-6 2012 Edition for this type of package.

Specifically, for exclusive use packages, Paragraph 573(a) of SSR-6 requires that the surface radiation level shall not exceed 2 mSv/hr unless the provisions in subpart (a) (i), (ii), and (iii) are met; then the surface radiation level shall not exceed 10 mSv/hr. Paragraph 573(b) requires that the maximum radiation level at the surface of the vehicle shall not exceed 2 mSv/hr and Paragraph 573(c) requires that the radiation level shall not exceed 0.1 mSv/hr at any point 2 m from the vertical planes represented by the outer lateral surfaces of the vehicle, or, if the load is transported in an open vehicle, at any point 2 m from the vertical planes projected from the outer edges of the vehicle. For Type B(U) packages, Paragraph 652 requires that the package meet the requirement in Paragraph 648 of SSR-6. Paragraph 648(b) states that under the tests for normal conditions of transport, the package cannot experience more than a 20% increase in the maximum radiation level. Paragraph 659(b)(i) of the SSR-6 requires that the package does not exceed 10 mSv/hr at 1 meter under accident conditions.

The revalidation request was limited to Gisete 4, Gisete 5 or Gisete 8 RTG placed in a special-purpose internal fitting, as described in Appendix 16 of the French certificate. The radionuclide is Sr-90 with a maximum activity of 897 TBq. The isotopic generators consist of a metal capsule locking the radioactive material, which forms the source block, and a primary shield where the source block is inserted. This source block is made from a tungsten alloy or a depleted uranium primary shield. There is also a thermal insulator and a closure system. The Gisete 4 and 5 contain a lead or cast iron secondary shield, and the Gisete 5 and 8 contain an outer case.

The TN-MTR packaging (including the lid) is in a cylindrical shape with a height of 2.008 meters and a diameter of 2.080 meters. The cavity has a height of 1080 mm and a diameter of 960 mm. The body is composed of lead, surrounded by thermal protection made of resin to protect from fire and is enveloped in two concentric stainless steel cylinders. It has a lid that is composed of lead surrounded by a stainless steel casing. The package includes a shock absorbing cover made of wood enclosed in stainless steel.

5.1 Source Term Sr-90 beta decays to Yttrium 90 (Y-90) which beta decays into Zirconium 90 (Zr-90), a stable nuclide. Because the half-life of Sr-90 is much longer than that of Y-90, and the time since loading is not specified, it is conservative to assume that the two would be in secular equilibrium. Although the beta emissions from the Sr-90 and Y-90 would be easily stopped by the package components, the beta particles interacting with the tungsten or uranium primary shield (high Z materials) would create bremsstrahlung photons that need to be considered when evaluating radiation levels. The maximum beta emitted from Sr-90 is 0.546 MeV with an average energy of 0.196 MeV and from Y-90 the maximum beta emitted is 2.284 MeV with an average energy of 0.935 MeV.

The applicant evaluated the bremsstrahlung source using the ORIGEN code which has pre-calculated bremsstrahlung estimates from electron interactions in UO2. The applicant submitted the gamma source term in the June 12, 2019 supplement to the application. The staff calculated the bremsstrahlung source as estimated from ORIGEN from SCALE 6.2.3. The staff used a decay time of 20 days. This was chosen to ensure Sr-90 and Y-90 were in secular equilibrium which occurs when decay time is greater than 7 daughter half-lives. The staff was

able to reproduce the applicants bremsstrahlung source in this manner giving it reasonable assurance it was calculated correctly with this method.

5.2 Package Model The applicant shows the dimensions it used for routine transport conditions for the TN-MTR package in Figure 4A-14.1 of Chapter 4A-Appendix 14 of the TN-MTR safety analysis report (Document No. DOS-18-011415-045). The staff compared the representation of the package body in the application to that of the drawings TN-MTR Packaging - Safety Drawing (General View), (Drawing Reference No. PLA-15-00166811-000). The staff found that the dimensions used by the applicant within the shielding model are equivalent to the dimensions in the licensing drawing. The applicant did not consider manufacturing tolerances as identified within the drawings. Although this is a non-conservative assumption, the staff found that the conservatism within the calculated radiation levels and in the source modeling discussed below would bound this uncertainty.

The staff reviewed the material densities the applicant used within the shielding evaluation within Table 4A-14.1 of Chapter 4A-Appendix 1 of the TN-MTR safety analysis report. The main shielding components are made of lead and steel. The staff found that the densities used for these components are the same as or conservative as compared to published values (PNNL-15870 Rev. 1, Compendium of Material Composition Data for Radiation Transport Modeling, Rev. 1, March 2011).

The applicant modeled the source as a point source adjacent to the bottom and the side of the TN-MTR package. This is conservative because the top of the package has the least amount of lead shield however there is additional steel and the shock absorbing cover which is mostly made of wood so does not provide as much shielding, however it does increase the distance to the detector for this surface by an additional 180 mm.

The staff performed a sensitivity study using the Microshield computer code and the gamma source provided by the applicant to determine the difference in dose rate when modeling the source at the top of the package versus the side and bottom of the package and found that it is conservative to model the source at the bottom and side of the package. The differences in the package geometry and shielding at the top reduce the radiation level, as compared to the side and bottom of the package by more than an order of magnitude. The staff used the American National Standards Institute (ANS)/ANSI-6.1.1-1977, Neutron and Gamma-Ray Flux-to-Dose Factors, flux-to-dose-rate conversion factors in its dose rate calculations. The staffs calculations also show that the dose rate for this package is very low and an order of magnitude less than the regulatory limit using a very conservative representation of the source (point source) and package.

The applicant neglects the Gisete RTG structural components in its model for calculating the radiation level. This approximation is conservative.

5.3 Accident Conditions Under accident conditions the applicant calculated the external radiation level assuming the resin is replaced by air, neglecting the cover, and implementing a reduction in lead height and thickness due to lead slump. The staff found that these considerations would adequately represent the package under accident conditions consistent with the drop and fire tests discussed in Chapters 1 and 2 of the application. The applicant modeled the source as a point

source in the top corner. The staff found that this is a conservative location for the source under accident conditions as it assumes the loss of the cover which reduces the distance to the detector, and there is lead slump in this location. The staff did not evaluate if the amount of lead slump assumed was appropriate and instead used its judgment that it is acceptable based on the fact that it is similar to values the staff has seen in other similar packages. The staff additionally based its finding on the assumption that the increase in dose rate due to the uncertainty in these modeling assumptions would not cause the package to exceed the regulatory radiation level due to the conservatism in the source term modeling (point source neglecting the RTG shielding material) and margin to the radiation level limit, under accident conditions calculated to be 2.07 mSv/hr (limit 10 mSv/hr required by SSR-6 paragraph 659(b)(i)).

5.4 Evaluation Method The applicant calculated the source term using the ORIGEN code from the SCALE6 package.

ORIGEN has pre-calculated bremsstrahlung contributions in UO2 and H2O. The Gisete RTG material is tungsten or uranium, and the uncertainty for assuming the bremsstrahlung medium is UO2 versus tungsten or uranium is unknown. However, the staff still found the bremsstrahlung approximation to be acceptable for this application considering the large margin to regulatory radiation level limits and the conservatisms within the modeling, such as not modeling the RTG components within the shielding evaluation.

The applicant evaluated the external radiation level of the package using the TRIPOLI 4.4 Code and the ENDF/B-VI cross section library. Although the staff has little experience with this code, it is internationally recognized and available through Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA/OECD). It is also distributed by the Radiation Safety Information Computational Center (RSICC) managed by Oak Ridge National Laboratory. The staff has reasonable assurance that the code is capable of performing the necessary radiation level calculations. In determining acceptability of this code for this evaluation, the staff considered its own calculations as well as the conservatism within the evaluation.

5.4.1 Flux-to-Dose Rate Conversion Factors The applicant uses the flux-to-dose-rate conversion factors for the ambient dose equivalent, H*(10), incorporated into TRIPOLI 4.4 from the International Commission on Radiological Protection, Publication 74 Conversion Coefficients for use in Radiological Protection against External Radiation, (ICRP-74). The staff accepts the flux-to-dose-rate conversion factors from ANS/ANSI-6.1.1-1977. The ambient dose equivalent in ICRP-74 was formulated as an operational quantity for area monitoring meaning that it can be measurable from a detector.

These conversion factors have not been evaluated by the NRC staff and the staff continues to accept the ANS/ANSI-6.1.1-1977 which more conservatively calculates the measurable dose rate. The staff compared the flux-to-dose-rate conversion factors between ANS/ANSI-6.1.1-1977 and that of H*(10) from ICRP-74 and found that the difference is as large as two orders of magnitude for lower energy photons (10 keV) and as much as 1.5 times higher for photons above 40 keV. The reason the conversion factors for ANS/ANSI-6.1.1-1977 are so much higher for lower energy photons is because the H*(10) conversion factors from ICRP-74 are evaluated assuming dose is deposited in a 10 mm International Commission on Radiation Units (ICRU) sphere. Lower energy photons deposit energy at a shallower depth. This is recognized by the guidance document of SSR-6, Specific Safety Guide Bo. SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (2012 Edition).

SSG-26 Paragraph 233.1 states that SSR-6 radiation levels are in terms of ambient dose equivalent for strongly penetrating radiation and directional dose equivalent for weakly penetrating radiation. General Safety Requirements Part 3 No. GSR Part 3, Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards, further defines the ambient dose equivalent as the dose equivalent produced at a recommended depth of 10 mm in the ICRU sphere and directional dose equivalent as the dose equivalent produced at a recommended depth of 0.07 mm in the ICRU sphere. GSR Part 3 defines weakly penetrating radiation to include photons of energy below about 12 keV, electrons of energy less than about 2 MeV, and massive charged particles such as protons and alpha particles.

The staff found that the flux-to-dose-rate conversion factors agree more between the ANS/ANSI-6.1.1-1977 and the directional dose equivalent H(0.07,0o) from ICRP-74 at energies below 30 keV.

Although the bremsstrahlung source provided by applicant does have the largest flux at the lowest energy (10-100 keV), these photons are statistically less likely to reach the detector through the shielding and those contributing to the dose rate will be the higher energy photons.

The energy of these photons is where the flux-to-dose rate conversion factors used by the applicant underpredict the dose rate by as much as a factor of 1.5. There is about a factor of 5 conservatism in the most limiting dose rate (accident conditions). Considering this and other conservatisms within the evaluation (such as neglecting the Gisete shielding) the staff found the use of the ambient dose equivalent flux-to-dose-rate conversion factors from ICRP-74 to be acceptable for this application.

5.4.2 Radiation Levels The applicant evaluated the radiation level under routine conditions for transport adjacent to where the source is located within package (top, bottom or side). The applicant found that the maximum radiation level on the surface is 0.0294 mSv/hr. This is below the limit of 2 mSv/hr from SSR-6 Paragraph 573(a) and 573(b). At 2 meters the applicant calculated the maximum radiation level of 0.000369 mSv/hr. This is below the limit of 0.1 mSv/hr from SSR-6 Paragraph 572(c). Under accident conditions, the applicant evaluated the radiation level 1 meter from the top of the package and the top side adjacent to the source and the lead slump and calculated the maximum radiation level of 2.07 mSv/hr. This is below the limit of 10 mSv/hr from SSR-6 Paragraph 659(b)(i).

In Appendix 4A.1 of the application, the applicant justifies how the package complies with the criterion in Paragraph 648(b) that requires that under normal conditions of transport the package cannot experience more than a 20% increase in the maximum radiation level. The staff reviewed the information in this appendix. This information shows that under the normal condition drop tests specified in Paragraph 722 of SSR-6 there is some deformation to the shock absorbing cover that would cause radiation levels to increase slightly due to the reduced distance to the detector. The applicant evaluated the difference in radiation levels and showed that they do not increase by more than 20%. The staff found the applicants evaluation demonstrates that the package meets the requirement of Paragraph 648(b) of SSR-6.

5.5 Conclusion As discussed in the above paragraphs, the staff has reasonable assurance that the TN-MTR package with the Gisete 4, Gisete 5 or Gisete 8 RTG placed in a special-purpose internal fitting, as described in Appendix 16 of the French certificate meets the requirements in Paragraphs 573, 648(b), 659(b)(i) in SSR-6. The staff recommends revalidation of French Certificate of Approval No. F/357/B(U)-96 Rev. Eaf for the TN-MTR package with these contents.

6.0 CRITICALITY EVALUATION

There is no fissile material in the package, therefore a criticality evaluation is not required.

7.0 OPERATING PROCEDURES EVALUATION Chapter 6A of the safety analysis report and the French approval certificate include sections on package acceptance, loading, unloading, and pre- and post-shipment requirements. The operating procedures have specific measures to be taken prior to each shipment, including ensuring the correct radial and axial wedges are used for the corresponding RTG, installing the package closures and confirming that the lid screws are properly torqued, leak testing the lid and containment penetration seals, and taking radiation measurements.

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM EVALUATION The acceptance test program includes: a) document examination; b) visual examination; c) leak testing; d) shielding integrity tests; e) thermal tests; f) packaging and basket operating tests; and g) packaging markings to ensure that the package is fabricated in accordance with the design approved in the French certificate. The maintenance program includes requirements for each shipment, after 15 cycles (or 3 years whichever is less), and after 60 cycles. The maintenance tests for each cycle, 15 cycles and 60 builds upon one another to ensure continued efficacy of the package design.

CONCLUSION Based on the statements and representations contained in the documents referenced above (see

SUMMARY

), the staff concludes that the Model No. TN-MTR package meets the requirements of International Atomic Energy Agency Regulations for the Safe Transport of Radioactive Material, IAEA Safety Standards Series, No. SSR-6, 2012 edition.

Issued with letter to R. Boyle, Department of Transportation, on ____________.