ML20227A029
| ML20227A029 | |
| Person / Time | |
|---|---|
| Site: | 07103096 |
| Issue date: | 09/10/2020 |
| From: | John Mckirgan Storage and Transportation Licensing Branch |
| To: | Boyle R US Dept of Transportation, Radioactive Materials Branch |
| NJDevaser NMSS/DFM/STL 415.5196 | |
| References | |
| EPID L-2019-LLA-0173 | |
| Download: ML20227A029 (19) | |
Text
R. Boyle UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 10, 2020 Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 400 Seventh Street, S.W.
Washington, D.C. 20590
SUBJECT:
RECOMMENDATION FOR REVALIDATION OF JAPANESE CERTIFICATE OF APPROVAL J/2027/AF-96 REVISION 0, FOR THE MODEL NO. RAJ-IIIS PACKAGE, (DOCKET NO. 71-3096)
Dear Mr. Boyle:
By letter dated June 20, 2019 (Agencywide Documents Access and Management System Accession No. ML19220A165), as supplemented on January 21, 2020 (ADAMS Accession No. ML20213B079) and April 6, 2020 (ADAMS Accession No. ML20216A744), the U.S. Department of Transportation requested that the U.S. Nuclear Regulatory Commission staff perform a review of the Japanese Certificate of Approval J/2027/AF-96, for the Model No. RAJ-IIIS transport package and make a recommendation concerning the revalidation of the package for import and export use.
Based upon our review, the statements and representations contained in the application, and for the reasons stated in the enclosed safety evaluation report, we recommend revalidation of the Japanese Certificate of Approval No. J/2027/AF-96, Revision 0, for the Model No. RAJ-IIIS package.
If you have any questions regarding this matter, please contact me or Nishka Devaser of my staff at (301) 415-5196.
Sincerely, Digitally signed by John B.
John B. McKirgan McKirgan Date: 2020.09.10 13:46:08 -04'00' John McKirgan, Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3096 EPID: L-2019-LLA-0173
Enclosures:
- 1. Safety Evaluation Report
(Transmittal letter and SER):
OFFICE: NMSS\DFM NMSS\DFM NMSS\DFM NMSS\DFM NMSS\DFM NAME: NDevaser SFigueroa JTapp ABarto JBorowsky DATE: 07/20/2020 07/21/2020 07/21/2020 07/27/2020 07/23/2020 OFFICE: NMSS\DFM NMSS\DFM NMSS\DFM NMSS\DFM NMSS\DFM NAME: TAhn PKoch CKenny ASotomayor-Rivera LCuadrado DATE: 07/21/2020 07/24/2020 07/27/2020 07/20/2020 08/07/2020 OFFICE: NMSS\DFM NMSS\DFM NMSS\DFM NMSS\DFM NAME: TBoyce RChang YDíaz Sanabria JMcKirgan DATE: 08/03/2020 07/31/2020 08/06/2020 9/10/2020 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3096 Model No. RAJ-IIIS Package Certificate of Approval No. J/2027/AF-96 Revision 0 Enclosure 1
Table of Contents Page
SUMMARY
................................................................................................................................ 1 1.0 GENERAL INFORMATION ............................................................................................ 1 1.1 Packaging Description .......................................................................................... 1 1.2 Drawings............................................................................................................... 2 2.0 STRUCTURAL EVALUATION ....................................................................................... 2 2.1 Structural Design and Requirements .................................................................... 2 2.2 General Requirements for all Packagings and Packages ...................................... 2 2.3 Requirements for Type A Packages...................................................................... 3 2.4 Tests for demonstrating ability to withstand normal conditions of transport ........... 4 2.5 Tests for demonstrating ability to withstand accident conditions of transport ......... 4 2.6 Evaluation Findings............................................................................................... 5 3.0 THERMAL EVALUATION .............................................................................................. 5 3.1 Packaging and Thermal Design Features ............................................................. 5 3.2 Material Properties and Component Specifications ............................................... 6 3.3 Thermal Design Limits of Package Materials and Components ............................ 6 3.4 Thermal Evaluation under General Conditions (i.e., Normal transport conditions) . 6 3.5 Thermal Evaluation under Special Test Conditions (i.e., Accident transport conditions) ...................................................................................................................... 7 3.6 Evaluation Findings............................................................................................... 8 4.0 CONTAINMENT EVALUATION ..................................................................................... 8 4.1 Description of the Containment System ................................................................ 8 4.2 Containment under General Conditions (i.e., Normal transport conditions) ........... 8 4.3 Containment under Special Test Conditions (i.e., Accident transport conditions) .. 9 4.4 Evaluation Findings............................................................................................... 9 5.0 SHIELDING .................................................................................................................... 9 5.1 Review Objective .................................................................................................. 9 5.2 Description of Shielding Design ............................................................................ 9 5.3 Radiation Source .................................................................................................10 5.4 Evaluation Findings..............................................................................................10
6.0 CRITICALITY EVALUATION
........................................................................................10 6.1 Review Objective .................................................................................................10 6.2 General ................................................................................................................10 6.3 Evaluation Findings..............................................................................................12 7.0 MATERIALS EVALUATION .........................................................................................12 7.1 General ................................................................................................................13 7.2 Materials Properties .............................................................................................13 7.3 Chemical and Galvanic Reactions .......................................................................13 7.4 Evaluation and Findings .......................................................................................14 8.0 QUALITY ASSURANCE ...............................................................................................14 8.1 Staffs Evaluation of the Quality Assurance Program ...........................................14 8.2 Evaluation Findings..............................................................................................14
9.0 REFERENCES
..............................................................................................................15 CONDITIONS ...........................................................................................................................15 CONCLUSION ..........................................................................................................................15
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3096 Model No. RAJ-IIIS Package Certificate of Approval No. J/2027/AF-96 Revision 0
SUMMARY
By letter dated June 20, 2019 (Agencywide Documents Access and Management System Accession No. ML19220A165), as supplemented on January 21, 2020 (ADAMS Accession No. ML20213B079) and April 6, 2020 (ADAMS Accession No. ML20216A744), the U.S. Department of Transportation (DOT) requested that the U.S. Nuclear Regulatory Commission (NRC) staff perform a review of the Japanese Certificate of Approval J/2027/AF-96, for the Model No.
RAJ-IIIS transport package 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 Daher-TLI in its application for the Model No. RAJ-IIIS package and its supplements against the regulatory requirements of the International Atomic Energy Agency (IAEA) Specific Safety Requirements No. SSR-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 Competent Authority (CoA)
J/2027/AF-96, Revision 0, Model No. RAJ-IIIS package, for shipment of the contents as described in Section 1.1.2, Contents, of this safety evaluation report (SER).
1.0 GENERAL INFORMATION 1.1 Packaging Description The Model No. RAJ-IIIS package is designed for the transport of unused Uranium fuel rods for the Static Experiment Critical Facility (STACY) in Japan called STACY fuel rods. The packaging is composed of two main components: an inner container that holds the fuel rods, which are contained within rod boxes, and an outer container that provides impact protection.
The inner container is a double-wall stainless steel box that incorporates alumina thermal insulator filling between the two walls. The outer container is constructed of a thin outer wall of stainless steel with balsa wood and resin-impregnated paper honeycomb inside to support the inner container. The outer container is also equipped with a shock absorbing suspension mechanism to protect the fuel rods from excessive vibration loads. The fuel rods are positioned within rod boxes, called protection cases.
The approximate dimensions and weights of the package are:
Overall width 73 cm Overall height 74 cm Overall length 507 cm Weight of packaging 920 kg Maximum weight of package, including contents 1490 kg
2 1.2 Drawings The packaging is constructed and assembled in accordance with the following Global Nuclear Fuel - Japan Drawings:
TTO-T06-047-01, Sheets 1-6, Rev. 0 Structure of Outer Container (RAJ-III)
TTO-T06-047-02, Sheets 1-6, Rev. 0 Structure of Inner Container (RAJ-III)
TTO-T06-047-03, Sheet 1, Rev. 0 Structure of Shock Absorber (RAJ-III)
The fuel rod box is constructed and assembled in accordance with Figures (A)-D.1 and (A)-D.2.
2.0 STRUCTURAL EVALUATION 2.1 Structural Design and Requirements The structural evaluation of the package is presented in Section (B)-A of the Safety Analysis Report. Section (B)-A.1 presents the structural design, including design standards, Section (B)-A.2 presents weights and center of gravity, Section (B)-A.4 addresses general standards for all packages, Section (B)-A.5 presents the evaluation of the package under normal conditions of transport (NCT), and Section (B)-A.6 presents the evaluation under accident conditions, including the 9-meter drop test (drop I), the puncture test (drop II), the thermal test, and immersion tests.
Evaluation of the RAJ-IIIS package against the relevant section of IAEA SSR-6 is provided below along with the relevant paragraph number of SSR-6.
2.2 General Requirements for all Packagings and Packages An adequate physical description of the package is provided in Section (A)-C of the application, along with the figures and tabular data for the packaging. The design meets the requirements of paragraph 607.
The structural design of the lifting system to be used to assemble, disassemble, and move the package is provided in Section (B)-A.4.4. The entire package is designed to be handled either by crane with wire-rope slings, or by forklift. Structural calculations have been provided for the various lifting conditions of the components of the package that have addressed induced bending stresses; however, the shear stresses induced through the body of the outer container have not been identified in the calculations for either mode of handling. This requirement has not been fully demonstrated by analysis to have been met. However, supplemental steel members are provided on the bottom of the outer container in these locations that will provide additional shear capacity. In addition, previous handling of the package to date has not resulted in any known failures or identified deficiencies (although the staff noted that handling experience to date does not quantify the margin against failure).
Therefore, the staff finds that the lifting system design is acceptable and the requirement in paragraph 608 has been met.
The outer container lid possesses sling fittings for the removal of the outer lid, and these are marked as, exclusive use for lid lifting only. This requirement (found in paragraph 609) has only been addressed through administrative procedures as an administrative control on the
3 user and not by physical configuration of the packaging. This may not be in conformance with the requirements, which specifies that the attachments must be designed to support the package mass or shall be removable or otherwise rendered incapable of being used during transport. The approval should be conditioned to specify that the lid lifting fixtures are rendered inoperable for package lifting during transport. See the Evaluations Findings section below and the Conditions section of this SER.
The stainless steel skin of the outer container, as described in Section C.2.2 of the application, provides for a free draining packaging that should not result in the retention of water. The requirement in paragraph 611 has been met.
In accordance with paragraph 613, the packaging has been designed to withstand the effects of accelerations and decelerations that could reasonably be expected to occur during the transport and handling.
However, vibration effects on bolted closures have not been addressed in detail in Section A.4.7. Figures (A)-C.2 and (A)-C.3 and Sections C.2.1 and C.2.2 identify that the closure of the inner and outer containers is accomplished with bolts. No information is provided that indicates any special considerations have been made regarding the assurance that these bolts will not loosen during the transportation conditions. It is noted that the closure bolts for the fuel rod protective case also does not appear to have had any special considerations made regarding bolt loosening during transport. It is recommended that a minimum and maximum bolt torque value be included in the package operating procedures.
The range of values should be based on the analytical method used in which the torque induced bolt tensile stresses were considered when analyzing the total bolt stresses under the accident conditions. It is noted in Section (D)- A.1.2(7) that after the outer lid is placed, the procedure is to tighten the bolts securely. A condition to address the adequate tightening of the closure bolts is in the Evaluations Findings section below and the Conditions section of this SER.
The design of the packaging for pressure and temperature is addressed in Sections A.4.2, A.4.6, and A.5.1 (paragraph 616). Computed stresses are within the appropriate allowable values and this requirement has been met. The air transport requirements of paragraphs 619, 620, and 621 have not been addressed. The approval should be conditioned to prohibit air transport. The condition is in the Evaluations Findings section below and the Conditions section of this SER.
2.3 Requirements for Type A Packages The physical dimensions of the packaging meet this requirement in paragraph 636.
The lead seal and seal pin configuration meet this requirement in paragraph 637.
The tie-down system is not part of the packaging, but wire-rope is used on the outer container configuration as well as an enclosing rack system for truck transport. The tie-down system and rack system used are compatible with the packaging design. This requirement (found in paragraph 638) has been met. The NRC staff notes that the lid lifting fixtures were not evaluated for package tie-down, and, therefore, the lid lifting fixtures must be rendered inoperable for package tie-down during transport.
The design and manufacturing techniques (referenced in paragraph 640) relative to the structural materials and their fabrication are in accordance with recognized standards. The
4 reference to the U.S. stainless steel known as 304L does not have physical properties (stress limits) that are consistent with the current material table values in the SAR, however, this type of stainless steel (304L) would be acceptable as a material of construction for this design.
The prototype tests were performed on a target that complies with the requirements in paragraph 717. The target consisted of a substantial reinforced concrete mat of 3 feet-9 inches thickness, covered with a 1-1/4-inch thick steel plate. The properties of the target for the drop II test complied with Paragraph 727(b).
2.4 Tests for demonstrating ability to withstand normal conditions of transport The water spray test was not performed since all materials exposed are not subject to degradation or soaking effects over short durations. The stacking effects were evaluated analytically as well as the penetration effects under normal conditions. Prototype free drop tests were performed. These requirements from paragraph 719 have been met.
The water spray test was not performed based on the outer container materials. This test condition from paragraph 720 did not apply.
The water spray test was not performed based on the outer container materials. This test condition from paragraph 721 did not apply.
Free drop tests from 1.2 meters were performed in three orientations, after a corner drop from 0.3 meter, for this package with a mass of less than 5000 kg. This requirement in paragraph 722 has been met.
This stacking test requirement was addressed by analysis, since the metallic structure could be realistically modeled and analyzed. This requirement in paragraph 723 has been met.
This penetration test requirement was addressed by analysis, since the metallic structure could be realistically modeled and analyzed. This requirement in paragraph 724 has been met.
2.5 Tests for demonstrating ability to withstand accident conditions of transport The mechanical free drop tests for accident conditions were performed prior to the thermal test. The requirement in paragraph 726 has been met.
For this packaging, the provisions of Paragraph 727(a) and (b) were applied requiring a free drop of 9 meters onto the proper target so as to suffer the maximum damage. There were two specimens tested after each had been subject to the 0.3-meter and 1.2-meter free drops required for normal conditions for fissile material packages. One specimen was subjected to a 9-meter free drop on the flat steel plate of the target so as to strike the corner of the bottom face, and then was dropped again to impact on the opposite end of the packaging with the end essentially horizontal. A second specimen was subjected to a 9-meter free drop on the flat steel plate of the target so as to strike the lid face in the horizontal position. This specimen was then subjected to the 1-meter free drop onto the rigid vertical bar 15 cm in diameter and 20 cm long. There was damage to the outer container, as well as the inner container, and some bending of fuel rods. Despite this damage, no rupture of the rods occurred. Therefore, the requirement in paragraph 727 has been met.
5 Water will enter this packaging and contact the exterior of the new fuel rods. As determined by analysis, the water pressures exerted on the new fuel cladding tube will be less than the pressure conditions caused under the general test conditions. Therefore, the structural analysis of this water immersion test is bounded by the pressure difference considered in the general test conditions. The requirement in paragraph 729 has been met.
2.6 Evaluation Findings
This licensing action is a revalidation of the Japanese certificate that was approved according to the IAEA SSR-6. Based on a review of the statements and representations in the application for the RAJ-IIIS package with Type A content, the staff finds that the applicant has not sufficiently demonstrated that the package meets the requirements of IAEA SSR-6 Paragraph 609, as the applicant has not demonstrated that the outer container lid sling fittings are designed to support the package mass or be removable or rendered incapable of being used during transport. The staff finds that the applicant has not sufficiently demonstrated that the package meets the requirements of IAEA SSR-6 Paragraph 613, as the applicant has not addressed the effects of vibration on the closure bolts. The staff finds that the applicant has not addressed the air transport requirements of Paragraphs 619, 620, and 621.
The staff recommends that the package be approved with the following conditions:
- The applicant renders the lid lifting fixtures inoperable for package lifting during transport in order to meet the requirements of IAEA SSR-6 Paragraph 609.
- The applicant addresses the adequate tightening of the closure bolts in order to meet the requirements of IAEA SSR-6 Paragraph 613.
- Air transport of the package is prohibited.
Based on a review of the statements and representations in the application, and with the conditions listed above, the staff concludes that there is reasonable assurance that the RAJ-IIIS package has sufficient structural capacity to meet the requirements of IAEA SSR-6.
3.0 THERMAL EVALUATION 3.1 Packaging and Thermal Design Features According to Section (A)-A and Section (A)-C of the application, the RAJ-IIIS is a Type A fissile package that consists of one inner container and an outer container constructed from stainless steel (Type 304/304L, per Table (A)-C.2). The outer container consists of a thin steel plate attached to angled pieces. The inner container is double-walled with two stainless steel sheets; alumina thermal insulation is placed between the double walls. Section B.1.1 indicated the thermal insulator protects the fuel rods from the fires heat input. Table (A)-C.2 indicated that pieces of lumber also act as an insulator for the package body and lid. The packages shock absorbers are constructed from balsa wood and honeycomb structure made from a resin impregnated craft paper.
Section (A)-D stated that content includes STACY fuel rods packed within two protection cases.
Section (D) indicated that the fuel rods, which are backfilled with helium, include uranium dioxide pellets (U-235, U-238 are the main nuclides) with enrichments no higher than 5%; the maximum radioactivity per package is 26 GBq; additional nuclide inventory of the uranium dioxide is presented in Section D.8 and Table (A)-D.7. As described in Section D.6 of the application, the low enriched uranium dioxide pellets are unused such that the decay heat is negligible.
6 The fuel rod includes a column of pellets surrounded by the zirconium alloy (density of 6.5 g/cm3) fuel cladding tube with upper and lower end plugs; the end plugs are hermetically sealed to the fuel tubes by welds. The zirconium alloy cladding, end plugs, and the seal welds form the containment boundary. The welds are inspected by X-ray or ultrasonic testing.
Section B.1.1 of the application indicated that fuse plugs are placed within the outer container and inner container. Specifically, the outer container includes two fuse plugs at the main body and two at the lid. Likewise, the inner container has four fuse plugs at the main body, two at the lid, and one at the end lid.
According to Section B.1.1 of the application, there is no supplementary system of cooling.
3.2 Material Properties and Component Specifications Table (B)-B.1 provided the density, specific heat, and thermal conductivity of the stainless steel and alumina thermal insulator that made up the packages inner and outer containers. Likewise, Section B.3 provided the density and thermal conductivity of the paper honeycomb shock absorber, alumina thermal insulator, and rubber gaskets. Section B.4.5 and Table (B)-B.3 listed the tensile strength, elongation, and yield strength (0.2%) of the zirconium alloy cladding.
Likewise, the mechanical properties of the steel and zirconium are in Table (B)- A.3.
3.3 Thermal Design Limits of Package Materials and Components Section B stated that the paper honeycomb shock absorber had an operational temperature limit of -185 deg C to 107 deg C. In addition, the alumina thermal insulator had an operational temperature limit up to 1300 deg C and the natural rubber gasket had an operational temperature limit between -50 deg C to 120 deg C. Section B.4.2 indicated that the package component with the lowest temperature limit is foam polyethylene, which has a limiting value of 90 deg C.
3.4 Thermal Evaluation under General Conditions (i.e., Normal transport conditions)
Section B Safety analysis of packages of nuclear fuel materials and others stated that the performance of the package under normal and accident conditions of transport were demonstrated by calculations and tests. According to Section B.4.1.1, the thermal models ambient boundary conditions included a 38 deg C ambient temperature and insolation of 800 W/m2 on the package outer containers top (ceiling) surface and 200 W/m2 on the outer containers wall surfaces. The thermal analysis relied on energy balances and correlations to determine the maximum temperature within the package would be 83 deg C; this value is less than the limiting value for steel, alumina insulation, and rubber gasket, as discussed above.
Finally, Section B.4.2 described maximum surface temperatures calculations using a 38 deg C ambient temperature and the insolation values noted above. Based on those assumptions, the maximum surface temperature was reported to be between 50 deg C and 55.5 deg C. Surface temperatures would be less than or equal to 50 deg C for packages in the shade (i.e., no insolation).
Section D.7 noted that the fuel rods are filled with helium and hermetically sealed such that the maximum back-fill internal pressure of the fuel rods is 0.5 MPa (absolute pressure). Section B.4.4 indicated that at 38 deg C ambient temperature, including insolation boundary conditions, and using Ideal Gas Law, the maximum pressure at normal conditions within the fuel rod would be 0.61 MPa (absolute). A calculation was performed to show a maximum normal operating pressure of 0.61 MPa (absolute) assuming the 38 deg C ambient temperature and the insolation
7 conditions described above. Section B.4.5 calculated a 3.97 MPa radial stress of the cladding tube containment boundary, which was below the 196 MPa yield strength; thus, the applicant concluded that the integrity of the fuel rods would be maintained. Likewise, Section B.5.5 of the application indicated that the radial stress of the rods from the above-mentioned internal pressures are less than the tensile strength of the zirconium alloys. Additional structural-related evaluations are presented in the Structural SER section.
Section B.4.3 noted that all package components can function when temperatures are at the cold normal conditions of -40 deg C.
3.5 Thermal Evaluation under Special Test Conditions (i.e., Accident transport conditions)
According to Section B.5, the applicant conducted a thermal test at special test conditions to determine the response of the package to a hypothetical accident fire. As described in Section B.5, the full-scale test package underwent successive drop tests so that damage to the package would be addressed during the thermal test. The package was then placed in a 38 deg C environment until equilibrium thermal conditions of the package were attained. The package was then exposed to an 800 deg C environment for 30 minutes (it is noted that air transport of this package is not allowed), including insolation, and then underwent the subsequent natural cooling process. Section B.5.1.2 indicated that the inner container included a dummy fuel bundle and dummy weight.
The furnace was a combustion gas convection type with burners on two sides. There were eight HS 40 type high-speed burners (400,000 kcal/h). The furnace dimensions were reported to be 3000 mm W x 200 mm H x 7000 mm L (note, the 200 mm H dimension appears to be a typographical error, recognizing that the package height greater than 200 mm). Twenty-five thermocouples were positioned throughout the test set-up to measure temperatures inside the furnace, outside outer container, inside of the outer container, outside of inner container, inside of the inner container, and the dummy fuel assembly.
The accuracy of the effective heating zone temperature was listed as +/- 20 deg C. It is noted that the maximum temperature of the outside outer container during the test was reported to approach 870 deg C, and therefore, the 20 deg C uncertainty is below the approximate 70 deg C margin above the regulatory 800 deg C accident thermal temperature.
Table (B)-B.4 of the application presented the maximum temperature and corresponding time within the tests transient. Likewise, Figures (B)-B.5 through (B)-B.10 presented the temperature profiles of the thermocouples recorded during the fires transient. The outside of the inner container had a reported temperature of 808.9 deg C. In addition, although the table reported contents to be 270.7 deg C, the text stated that the fuel rods were assumed to be 401 deg C, which was the reported inside temperature of the inner container.
Section 3.2.5 noted that packing materials in the package include plastic corrugated sheets, plastic caps, plastic bags, and plastic sheets. As mentioned previously, lumber, balsa wood and resin impregnated craft paper are also components of a package. Section B.5.3 and B.5.6 indicated that combustion of the paper honeycomb and melting of packing material were observed as a result of the fire test condition. However, Section B.5.6 presented no issues with the structural integrity of the fuel rods and indicated the effects of the degradation of the above-mentioned materials were considered when performing the criticality analysis.
8 Section B.5.4 indicated that at a cavity temperature of 446 deg C at the fire accident conditions, the maximum pressure within the fuel rod would be 1.13 MPa (gauge). Section B.5.6 indicated that the 1.13 MPa (gauge) pressure in the fuel rod would result in a radial stress of 8.78 MPa, which is below the 115 MPa tensile strength of the zirconium alloy. Additional structural-related evaluations are presented in the Structural SER section.
3.6 Evaluation Findings
This licensing action is a revalidation of the Japanese certificate that was approved according to the IAEA SSR-6. Based on a review of the relevant portions of the Japanese certificate and the representations in the application, the staff has reasonable assurance that the RAJ-IIIS package with STACY fuel rod Type A content meets the thermal requirements of IAEA SSR-6.
4.0 CONTAINMENT EVALUATION 4.1 Description of the Containment System The RAJ-IIIS is a Type A fissile package that consists of one inner container and an outer container constructed from stainless steel. As described in the safety analysis report (SAR)
Section (A)D.2.4, the containment boundary of this package is the fuel rods. The fuel rod is composed of a column of uranium dioxide pellets loaded into a Zirconium alloy fuel cladding tube with upper and lower end plugs. The cladding tube clearance is filled with helium gas and the cladding tube is sealed hermetically to welded upper end plug. The zirconium alloy cladding, end plugs, and the seal welds form the containment boundary. The applicant states in SAR Section (A)D.2.4 that all welds of the fuel rods are certified for integrity by X-ray inspection or ultrasonic testing. In response to RAI 4-1, the applicant explained the X-ray inspection acceptance criterion is more than 2 sigma, where sigma means 0.54 mm. A description of the main elements and dimensions of the fuel rods are shown in Tables (A)-D.3 and (A)-D.4 and a picture of the containment boundary is shown in Figure (A)-D.3. The applicant states in Section D.7 that the helium back-fill internal pressure of the fuel rods is max 0.5 MPa (absolute pressure).
4.2 Containment under General Conditions (i.e., Normal transport conditions)
Under normal conditions, the package was evaluated by analysis with a stacking test, penetration test, water pray test, and thermal test as described in the structural and thermal sections of the SAR. A prototype free drop test was also conducted.
The applicant confirmed the integrity of fuel rods by analyzing rising internal pressures of the fuel rods. As described in the structural and thermal evaluations of this safety evaluation report (SER), the maximum pressure within the fuel rod at normal conditions would be 0.61 MPa (absolute), due to heat-input of solar radiation. In the structural and thermal analyses, the applicant indicated that the radial stress of the rods from the above-mentioned internal pressures are less than the tensile strength of the zirconium alloys of the fuel rod. This is shown in SAR paragraph (B)-A.5.1.3 and evaluated in the structural section of this SER. Thus, the applicant states the integrity of fuel rods will be maintained under normal conditions.
Accordingly, staff finds that the containment of the RAJ-IIIS package will not be negatively impacted by the structural and thermal tests under normal transport conditions.
9 4.3 Containment under Special Test Conditions (i.e., Accident transport conditions)
Per SAR Section C.4, a series of prototype structural drop tests were conducted under accident conditions. No demonstration tests were conducted on the STACY fuel rods transported in the RAJ-IIIS package. The demonstration tests were conducted on fuel rods for a BWR assembly, and analysis of the fuel rods transported in the RAJ-IIIS package were conducted based on the test results, as discussed in the structural and material sections of this SER.
The applicant conducted a thermal test to determine the response of the package to the fire accident conditions. As described in the thermal section of this evaluation, the applicant conservatively assumed the fuel rods were 401 deg C. Under these conditions the internal pressure of the fuel rods at this maximum fuel rod temperature is 1.23 MPa (absolute). The applicant indicates that the radial stress of the rods from the internal pressures at accident conditions are less than the tensile strength of the zirconium alloys. Additionally, in SAR Section (B)-C.2.3 the applicant performed a helium leakage rate test of the fuel rods. The results shown in SAR Table (B)-C.1 confirm there was no leakage of radioactive material.
Accordingly, staff finds that the applicant has confirmed the integrity of the fuel rods under accident transport conditions, and the structural and thermal tests do not have a negative impact on containment of the fuel rods.
4.4 Evaluation Findings
This licensing action is a revalidation of the Japanese certificate which was approved according to IAEA SSR-6. Based on the review of the statements and representations in the application for the RAJ-IIIS package with Type A content, the staff finds that the applicant adequately described and evaluated the containment system of the RAJ-IIIS package, and it is acceptable.
Therefore, the staff concludes that there is reasonable assurance that the RAJ-IIIS package meets the containment requirements of IAEA SSR-6.
5.0 SHIELDING 5.1 Review Objective The objective of the shielding review is to verify that the Certificate J/2027/AF-96 for the RAJ-IIIS Package is acceptable for revalidation and complies with the design requirements for a package containing Non-Irradiated Nuclear Fuel Rod (Uranium Dioxide), specified in IAEA SSR-6.
5.2 Description of Shielding Design 5.2.1 Shielding Design The RAJ-IIIS is based on the existing RAJ-III package (DOT CAC USA/0595/AF-96). The RAJ-IIIS is designed to ship unirradiated, uranium dioxide fuel rods, with enrichments up to 5 wt.% U-235, for the STACY experimental facility at the Japan Atomic Energy Agency. As described in Section A of the SAR, the RAJ-IIIS is comprised of one inner container and one outer container, both made of stainless steel.
10 5.3 Radiation Source The requested validation is for unirradiated, uranium dioxide fuel rod contents, with an enrichment up to 5 wt.% U-235, limited to a Type A quantity of radioactive material.
The packaging contains the STACY fuel rods. One package can be loaded to a maximum 260 rods, 216 kg of UO2 pellets, whose enrichment is 5.0% or below. Specification of the radioactive contents are presented in Table 2 of the CoA. The maximum total activity of the fuel content is 26 GBq. Activity for the isotopes identified in the fuel content is as follows:
- U-232: maximum 1.58x10-2 GBq
- U-234: maximum 2.21x101 GBq
- U-235: maximum 7.64x10-1 GBq
- U-236: 1.15x10-1 GBq
- U-238: 2.26x100 GBq The Criticality Safety Index (CSI) of the package is 0.25 and the TI is 0.5, as noted in certificate J/2027/AF-96 Section 7 and Safety Analysis Section A of the application.
The staff verified and found that all the isotopes, especially U-232, are under the regulatory limits for A2 values specified in Section IV, Table 2 of the SSR-6, Edition 2012.
5.4 Evaluation Findings
The staff determined that the maximum quantity of Uranium nuclides within Table 2 of the certificate are within the A2 limits in Table 2 of the SSR-6, Edition 2012. Since this is a Type AF package, there is no need for a shielding evaluation and the staff finds it acceptable for revalidation.
6.0 CRITICALITY EVALUATION
6.1 Review Objective The applicant requested U.S. revalidation of the CoA for the Model No. RAJ-IIIS package for transportation of unirradiated loose UO2 rods enriched up to 5.0 weight percent (uranium-235) 235 U to the requirements of IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material. The packaging consists of outer and inner containers as shown in Figures (B)-E.1 and (B)-E.2 of the (safety analysis report) SAR, respectively. Each container consists of stainless steel shells and structural reinforcement, with balsa and paper honeycomb shock absorbers in the outer container and alumina silicate thermal insulator between the inner and outer shells of the inner container. The inner container has two long channels for placement of the contents.
6.2 General The requested contents consist of unirradiated loose zirconium alloy clad UO2 rods enriched up to 5.0 weight percent 235U. The rods have a maximum active fuel length, a maximum UO2 pellet diameter, and a minimum clad thickness of 0.054 cm. The rods are packaged in two axial locations within a protective case, shown in Figure (A)-D.2 of the SAR, with two groups of 65 rods, and the groups are separated by a central stainless steel block. Two protective cases are loaded into the inner container, one case in each of the containers two channels. Each rod is wrapped in a thin polyethylene sleeve, and additional plastic shoring material may be included
11 between rods. Fewer than 260 rods may be shipped, in which case dummy stainless steel rods must be included with the loaded fuel rods, so that the rods remain tightly packed in the protective case.
The applicant modeled the rods with an active length of 339.4 cm, conservatively neglecting the central stainless steel block in the protective case and increasing the amount of fissile material.
The applicant modeled the maximum pellet diameter and minimum clad thickness, and neglected the rod-clad gap, effectively modeling the cladding with a smaller outer diameter.
These modeling parameters are conservative in that they reduce the displacement of moderator within the modeled fuel lattice. The applicant also modeled varying amounts of water or polyethylene within the protective case, since polyethylene packing material may be present, and varied the pitch and number of rods to find the optimum moderation ratio. The applicant considered the rods moderated by varying densities of water, as well as by full density polyethylene filling the protective case. The applicants analyses showed that full density polyethylene resulted in the maximum system (k-effective) keff in all cases, and modeled the package this way in subsequent analyses. Therefore, the amount of polyethylene that may be present within the protective case is unlimited. The materials properties for the fuel contents and packaging materials are given in Tables (B)-E.4 and (B)-E.5 of the SAR. The staff reviewed the applicants contents model and materials properties used in the criticality evaluation and finds them to be acceptable and conservative for demonstrating maximum package reactivity.
The applicant evaluated single packages in isolation under NCT and HAC, as required by paragraphs 680 - 682 of IAEA SSR-6. The HAC model bounds the consideration of a single package under routine conditions of transport, required to be evaluated by paragraph 682(a).
The applicant considered varying densities of water separately within the inner and outer packaging and assumed either water or polyethylene moderation inside the protective case.
The applicant considered full water reflection of 20 cm surrounding the package under all conditions, and considered the maximum deformation experienced under the tests of paragraph 685(b) of IAEA SSR-6 under HAC. For the HAC model, the applicant considered melting or burning away of the impact absorbers within the outer packaging and replacement by water. The resulting keff plus three times the Monte Carlo calculation uncertainty (keff + 3) for a single package under NCT and HAC are 0.558 and 0.628, respectively. These calculated keff values are significantly below 0.95 and are therefore acceptable.
The applicant evaluated arrays of packages under NCT and HAC and determined a package CSI, as required by IAEA SSR-6 paragraphs 684 - 686. The applicant modeled both arrays reflected with 20 cm of full density water and varied the moderation conditions similar to the single package analysis. For NCT, the applicant evaluated an 18 x 20 x 3 array of 1080 packages and calculated a maximum keff + 3 that is acceptable. For HAC, the applicant evaluated an 18 x 12 x 2 array of 432 packages and calculated a maximum keff + 3 that is acceptable. Both of these calculated keff values are significantly below 0.95 and are therefore acceptable.
Using the number of packages in an array demonstrated to be subcritical under NCT and HAC, the applicant determined a CSI according to the requirements of IAEA SSR-6 paragraph 686.
The number N determined was the same under both NCT and HAC, so the resulting CSI is 0.25. Although NRC regulations require that the CSI be rounded up to the nearest tenth, the advisory material for IAEA SSR-6 in IAEA Specific Safety Guide No. 26 (SSG-26), Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, allows use of the exact value of the CSI in paragraph 686.3. Therefore, the staff accepts the applicants calculated CSI of 0.25.
12 For all criticality calculations, the applicant used the CSAS5 sequence of the SCALE 6 code package, with the KENO V.a Monte Carlo radiation transport code and the 238-group ENDF/B-VI neutron cross section library, using the cell-weighted cross sections option to model the fuel, clad, and moderator inside the protective case. The SCALE code system is a standard in the nuclear industry for performing Monte Carlo criticality safety and radiation shielding calculations and is therefore acceptable for this application. The applicant performed a benchmark test of the code and cross section library in Section E.5 of the SAR. The applicant modeled four critical experiments with low enriched UO2 rods in water and determined a bias of 0.007 in keff.
The applicants benchmark analysis contained several deficiencies, when compared to NRC guidance on code benchmarking contained in NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages. These deficiencies include: 1) dissimilarities between critical experiments and the modeled system, including enrichment and clad material; 2) no demonstration that the model is within the range of applicability of the selected experiments for key system parameters (e.g., enrichment, hydrogen to fissile ratio, energy of the average lethargy causing fission); 3) insufficient number of experiments modeled to determine code bias and bias uncertainty using normal statistics; 4) no analysis of code bias trends as a function of key system parameters; and 5) the calculated bias and bias uncertainty are not included in the final keff results. However, the staff accepts the applicants final calculated keff results for the following reasons: 1) there is significant margin between the applicants maximum calculated keff and the 0.95 acceptance criterion (>0.09 in keff); 2) the applicant used a code (SCALE 6) that is a standard in the industry for performing criticality calculations, and it is unlikely that a code bias and bias uncertainty calculated for water moderated low-enriched UO2 fuel rods in a package would be greater than the calculated margin of subcriticality; and 3) the staffs independent analysis (discussed below), using a more modern version of SCALE with more modern cross section data, confirmed the applicants calculated keff values.
The staff performed independent confirmatory analyses of the RAJ-IIIS package using the CSAS6 sequence of the SCALE 6.2.3 code system, with KENO VI and the continuous-energy ENDF/B-VII.1 cross section library. The staff modeled the single package and arrays of packages using assumptions and modeling parameters similar to those used by the applicant.
The staffs independent evaluation resulted in keff values that were similar to, or bounded by, the applicants results.
6.3 Evaluation Findings
The staff reviewed the CoA for the Model No. RAJ-IIIS package, as well as the applicants initial assumptions, model configurations, analyses, and results in the SAR. The staff finds that the applicant has identified the most reactive configuration of the Model No. RAJ-IIIS package with the requested contents, and that the criticality results are conservative. Therefore, the staff finds with reasonable assurance that the package, with the requested contents, will meet the criticality safety requirements of IAEA SSR-6.
7.0 MATERIALS EVALUATION The purpose of the materials evaluation is to verify that the performance of the materials used to fabricate the package meets the regulatory requirements of IAEA SSR-6.
13 7.1 General Fuel Specifications, Cask Design/Materials, and Environmental Conditions Fuel Specifications - Type A fissionable package.
Cask Design/Materials - outer container (stainless steel, Balsa, paper honeycomb and natural rubber) and inner container (stainless steel, alumina thermal insulator, natural rubber).
Environmental Conditions - no inerting, radiation and temperature of Type A package for NCT and HAC.
Engineering Drawing Package Description_2A. 201_Rev.1 complies with the IAEA SSR-6 requirements include: 566 (b), 612, 614, and 616; Type A Package (635 - 651); Containing Fissile Material (673 - 686).
Based on compliance with the definitions and the IAEA SSR-6 requirements, the staff finds that the general considerations of the materials described above are acceptable.
7.2 Materials Properties The staff reviewed the stainless steel properties provided in SAR Table (B)-A.3 and verified that they comply with the cited ASTM standards (and accordingly for IAEA SSR-6). The staff also found that the mechanical properties for other materials (aluminum - if included as in 2007, alumina, foam natural rubber, lumber, Balsa) are acceptable, based on drop test results and data available in the technical literature.
The staff also verified that the allowable limits for temperature and pressure for the polymeric package materials are acceptable (Thermal_04 B.BC Rev.1). The packages radiation level of a few rad is well below limiting level for polymer seals (e.g., 107 rad for Parker O-Rings). Welding design and specifications comply with acceptance criteria of the nondestructive x-ray (RT) and ultrasonic (UT) examinations to check for defects or internal integrity of the welds after welding.
Materials property at low temperature to - 40 °C are included in the SAR, although the staff notes that there are no structural materials which can become brittle (e.g., ferritic steel) at low temperature. The staff also finds that the materials tested rod is representative of the actual fuel rods to be transported.
From these staffs assessments, the staff determines that the materials properties used are acceptable.
7.3 Chemical and Galvanic Reactions In a thermal test under an atmospheric environment of 800 °C for 30 minutes, the criteria for leakage of radioactive material are met, with no leakage. Similarly, under drop conditions, the criteria are met. Under HAC conditions, the material properties are as expected.
The package has no dissimilar metal contacts. Only stainless steel and zirconium alloys are used, and the zirconium alloy fuel tubes are separated from the stainless steel protection case with plastic sheets. Therefore, galvanic corrosion is not expected to occur. With the open system, given the range of operation temperature and short operation time of transportation, galvanic reaction rate will be negligible based on the staffs research experience and corrosion handbooks.
14 With the open system, any radiolysis of moisture will not accumulate hydrogen if produced. The amount of radiolysis-induced hydrogen will be minimal for a Type A fissionable package.
Based on the application, the staff determines that the design of the package adequately prevents significant chemical and galvanic reactions, and, therefore, the staff finds it acceptable.
7.4 Evaluation and Findings Based on a review of the statements and representations in the application for the RAJ-IIIS package with Type A content, the staff finds that the applicant adequately described and evaluated the materials performance of the RAJ-IIIS package, and they are acceptable.
Therefore, the staff concludes that there is reasonable assurance that the RAJ-IIIS package meets the containment requirements of IAEA SSR-6.
8.0 QUALITY ASSURANCE The purpose of the quality assurance (QA) review is to verify that the packaging QA program meets the requirements of the IAEA SSR-6. The staff reviewed the description of the QA program for the Model No. RAJ-IIIS package against the standards in the IAEA SSR-6.
8.1 Staffs Evaluation of the Quality Assurance Program The applicant developed and described a QA program for activities associated with transportation packagings for nuclear fuel materials. Those activities include design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use. The applicant described the QA organizations independence from other branches in the organization, which includes those responsible for product cost and schedule. The applicants description of the QA program [i.e., management system in IAEA SSR-6] meets the applicable requirements of IAEA SSR-6 and is based on ISO 9001:2000 and 10 CFR Part 50, Appendix B, among other standards. The staff finds the QA program description acceptable, since it allows implementation of the associated QA program for the design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use of the Model No. RAJ-IIIS transportation package.
The staff finds, with reasonable assurance, that the QA program for the RAJ-IIIS transportation packaging:
- a. meets the requirements in IAEA SSR-6, and
- b. encompasses design controls, materials and services procurement controls, records and document controls, fabrication and maintenance controls, nonconformance and corrective actions controls, an audit program, and operations or programs controls, as appropriate, adequate to ensure that the package will allow safe transport of the radioactive material authorized in this approval.
8.2 Evaluation Findings
Based on review of the statements and representations in the Model No. RAJ-IIIS package application and as discussed in this SER section, the staff has reasonable assurance that the RAJ-IIIS package meets the requirements in IAEA SSR-6.
15
9.0 REFERENCES
(DOT, 2019) U.S. DOT, U.S. Department of Transportation, "Transmittal of Japanese Certificate of Competent Authority J/2027/AF-96, Rev. 0 for the RAJ-IIIS package for Review," ML19220A165.
(DOT, 2020a) U.S. DOT, U.S. Department of Transportation, "Response to NRC Request for Supplementary Information for United States Validation of Certificate J/2027/AF-96 for the RAJ-IIIS Package,"
(DOT, 2020b) U.S. DOT, U.S. Department of Transportation, "Response to NRC Request for Additional Information for United States Validation of Certificate J/2027/AF-96 for the RAJ-IIIS Package,"
(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 SSR-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.
CONDITIONS Based on the staffs review, the staff recommends that DOT revalidate the certificate for import and export use, with the following additional conditions:
- 1. Transport by air is not authorized.
- 2. Lid lifting fixtures must be rendered incapable of being used for package lifting and tie-down during transport.
- 3. Package closure bolts must be adequately secured and torqued to prevent loosening during transport. Minimum and maximum torque values should be included in package operating procedures.
CONCLUSION Based on the statements and representations presented in the Safety Analysis Report and supplemental information, the staff agrees that the package meets the standards in IAEA Safety Standards SSR-6, 2012 Edition. The staff recommends that DOT revalidate Japanese Certificate of Approval No. J/2027/AF-96, Rev. No. 0, for import and export use, with the conditions listed above.
Issued with letter to R. Boyle, U. S. Department of Transportation.