ML23191A148
| ML23191A148 | |
| Person / Time | |
|---|---|
| Site: | 07109309 |
| Issue date: | 07/14/2023 |
| From: | Storage and Transportation Licensing Branch |
| To: | Global Nuclear Fuel |
| Shared Package | |
| ML23191A147 | List: |
| References | |
| EPID L-2022-LLA-0015, CoC No. 9309 | |
| Download: ML23191A148 (6) | |
Text
SAFETY EVALUATION REPORT Docket No. 71-9309 Model No. RAJ-II Package Certificate of Compliance No. 9309 Revision No. 13
SUMMARY
By letter dated September 23, 2022 (Agencywide Documents Access and Management System
[ADAMS] Accession No. ML22266A187), Global Nuclear Fuel - Americas, L.L.C. (GNF-A) requested a revision and renewal of NRC Certificate of Compliance (CoC) USA/9309/B(U)F-96 to allow for the transportation of GNF 10x10 fuel assemblies with enrichments up to 8 weight percent (wt%) 235U.
On June 12, 2023, GNF-A subsequently submitted the RAJ-II package general arrangement drawings as identified in section 1.3.1.1 of revision 11 of the RAJ-II safety analysis report (SAR),
NEDE-33869P, as well as changed pages to revision 11 of the RAJ-II SAR, NEDE-33869P to include a few clarifications requested by the NRC staff (ADAMS No. ML23163A210).
Based on the statements and representations in the application, and the conditions listed in the CoC, the U.S. Nuclear Regulatory Commission staff (the staff) concludes that the package meets the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.
EVALUATION
- 1. GENERAL INFORMATION
The Model No. RAJ-II package is used to transport unirradiated boiling water reactor (BWR) fuel assemblies, and BWR or CANDU or PWR fuel rods, containing Type B fissile material. The fissile material can be in the form of uranium dioxide or uranium carbide enriched up to 5.0 wt%
235U, or uranium dioxide up to 8.0 wt% 235U.
The contents consist of enriched commercial grade uranium or enriched slightly contaminated uranium with trace quantities limits as defined in ASTM C996-20, uranium oxide or uranium carbide fuel rods enriched to 5.0 wt% 235U for LEU or uranium oxide fuel rods enriched to no more than 8.0 wt% 235U for LEU+, with limits specified in table 1-2 and table 1-3 of the application. For the limits of table 1-3, 238U is assumed as being the remainder of the uranium concentration.
The packaging is constructed and assembled in accordance with the following Drawing Nos.:
- Outer/Inner Container Assembly Licensing Drawing No. 105E3737, Rev. 9
- Outer Container Main Body Assembly Licensing Drawing No. 105E3738, Sheet 1, Rev. 11
- Outer Container Main Body Assembly Licensing Drawing No. 105E3738, Sheets 2 and 3, Rev. 10.
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- Outer Container Fixture Assembly Licensing Drawing No. 105E3739, Rev. 6
- Outer Container Fixture Assembly Installation Licensing Drawing No. 105E3740, Rev. 7
- Outer Container Shock Absorber Assembly Licensing Drawing No. 105E3741, Rev.
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- Outer Container Bolster Assembly Licensing Drawing No. 105E3742, Rev. 5
- Outer Container Lid Assembly Licensing Drawing No. 105E3743, Rev. 7
- Outer Container Marking Licensing Drawing No. 105E3744, Rev. 8
- Inner Container Main Body Assembly Licensing Drawing No. 105E3745, Rev. 11
- Inner Container Parts Assembly Licensing Drawing No. 105E3746, Rev. 4
- Inner Container Lid Assembly Licensing Drawing No. 105E3747, Rev. 6
- Inner Container End Lid Assembly Licensing Drawing No. 105E3748, Rev. 4
- Inner Container Marking Licensing Drawing No. 105E3749, Rev. 8
- RAJ-II Protective Case Licensing Drawing No. 105E3773, Rev. 2
- Shipping Container Loose Fuel Rods Drawing No. 0028B98, Rev. 2
The Outer Container Shock Absorber Assy Licensing Drawing, 105E3741, applies to both the aluminum and paper honeycomb shock absorber. The aluminum honeycomb material meets the same mechanical properties as the paper material detailed in table 2-3 of the application.
2.0 STRUCTURAL EVALUATION
There is no change to the structural evaluation of the package.
3.0 THERMAL EVALUATION
3.1 Review Objective
The objective of the amendment review was to verify that the changes to the RAJ-II package thermal design were adequately described and evaluated under normal conditions of transport (NCT) and hypothetical accident conditions (HAC), as required per 10 CFR Part 71. As noted in SAR section Revision 11 Detailed Revision Summary (page iv), the amendments SAR changes reflected an increase from 5 wt% 235U (LEU) to 8 wt% 235U (LEU+) enrichment of GNF 10x10 BWR fuel assemblies and 10x10 BWR fuel rods as content and the incorporation of aluminum as an optional honeycomb construction material for the packages shock absorbers.
Regulations applicable to the thermal review include 10 CFR 71.31, 71.33, 71.35, 71.43, and 71.51.
3.2 General Considerations
There were no significant changes to the package thermal models described in the application.
As noted in SAR section 1.2.2.7, the decay heat associated with the 8.0 wt% 235U unirradiated content is essentially the same as the decay heat for 5 wt% enrichment fuel of earlier amendments; the decay heats for both enrichments were less than 0.25 W.
With respect to the RAJ-II packaging, SAR section Revision 11 Detailed Revision Summary and SAR section 1.2.1 indicated the packages honeycomb shock absorber material as either paper or aluminum. Aluminum has a higher thermal conductivity than paper and, therefore, previous amendment NCT thermal analyses using paper honeycomb thermal properties would be bounding.
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Regarding the thermal HAC, SAR section 2.12.2 indicated that package test units with paper honeycomb shock absorbers underwent certification fire testing; SAR section 3.1.3 and table 3-7 indicated that paper honeycomb is the bounding shock absorber material for the thermal analysis due to its combustive properties during HAC. Finally, SAR section 2.2.2 indicated that inspections, which include checking for corrosion, are performed prior to loading fuel.
3.3 Evaluation Findings
Based on a review of the thermal-related sections of the application, the staff concludes that the thermal design has been adequately described and evaluated and has reasonable assurance that the package meets the thermal requirements of 10 CFR Part 71.
4.0 CONTAINMENT EVALUATION
4.1 Review Objective
The objective of the amendment review was to verify that changes to the Type B RAJ-II package containment design were adequately described and evaluated under NCT and HAC conditions, as required per 10 CFR Part 71. There were no changes to the containment boundary or its testing as part of this amendment. Rather, regarding containment issues, the applicants revised safety analysis report reflected the increase from 5 wt% 235U (LEU) to 8 wt%
235U (LEU+) enrichment of GNF 10x10 BWR fuel assemblies and 10x10 BWR fuel rods as content.
Regulations applicable to the containment review include 10 CFR 71.31, 71.33, 71.35, 71.43, and 71.51.
4.2 General Considerations
SAR table 6-1, table 6-2, and table 6-3 specified the fuel loadings for the RAJ-II package. These three tables indicated that LEU+ GNF 10x10 BWR fuel assemblies and LEU+ 10x10 BWR fuel rods can be fueled with up to 8 wt% 235U enrichment.
The application included edits to SAR chapter 4 Containment regarding the addition of the 8.0 wt % 235U enriched LEU+ contents. The maximum concentrations of the radionuclides in the 5 wt% 235U and 8 wt% 235U content were provided in SAR table 4-1 and table 4-2, respectively.
These tables showed that most radionuclide concentrations did not change between LEU and LEU+ content; as a result, differences in total activity and effective A2 reported in SAR table 4-3 and table 4-4 were negligible.
Correspondingly, the NCT and HAC release calculations provided in SAR section 4.5.1.3 and section 4.5.1.4 showed essentially no change in the individual fuel rod allowable leakage rates between the LEU and LEU+ contents.
The calculated NCT and HAC individual fuel rod allowable leakage rates continued to be bounded by the qualification leak rate associated with certification test unit fuel assemblies described in SAR section 2.12.2.4.
In addition, SAR section 4.4 and section 8.1.4 noted each fuel rod is leak tested during fabrication to a leaktight acceptance criteria (less than 10-7 ref cc/sec, per ANSI N14.5).
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Based on the above, conclusions related to radionuclide release remain consistent with the previously certified package.
4.3 Evaluation Findings
Based on a review of the containment-related sections of the application, the staff concludes that the containment design has been adequately described and evaluated and has reasonable assurance that the package meets the containment requirements of 10 CFR Part 71.
6.0 CRITICALITY EVALUATION
The applicant requested to modify the Certificate of Compliance (CoC) for the RAJ-II BWR fresh fuel package to authorize transport of uranium oxide (UO2) fuel with 8.0 wt% 235U, referred to as LEU Plus (LEU+) fuel, analyzed within the GNF 10x10 fuel assembly.
The applicant evaluated the Model No. RAJ-II package containing two GNF 10x10 LEU+ fuel assemblies with two different fuel rod arrangements. The first case consisted of two fuel assemblies, each with 22, 2.0 wt% gadolinium oxide (Gd2O3) fuel rods and 16 part-length fuel rods. The second case consisted of 10x10 BWR loose fuel rods in each side of the package.
Each case was evaluated with worst case fuel parameters determined from previous analyses for 5.0 wt% fuel.
The normal conditions of transport (NCT) models for the LEU and LEU+ assemblies are identical except that a polyethylene thickness of 1.14 cm is used for the LEU+ assemblies compared to a thickness of 1.28 cm for LEU. Although this is not the worst-case scenario, the applicant shows in figure 6-42 of the application that throughout all ranges of polyethylene thickness the system reactivity is below the upper subcritical limit (USL) determined in the applicants benchmarking analysis.
This model for the RAJ-II GNF 10x10 was used in the NCT single package, HAC single package, NCT package array (21x3x24), and HAC package array (8x8) calculations.
To determine the criticality safety of an array of loose fuel rods for the 10x10 LEU+ fuel type, the applicant performed studies of moderator location, moderator density, and fuel rod pitch. For these models, the inner container of each package is filled with water while the outer container contains no water to facilitate neutron leakage between the packages. The applicant varied the inner container water density and found the worse-case density to be full density for the LEU+
fuel. The applicant varied the fuel rod pitch within the model and determined the pitch with the highest reactivity. Prior analyses of lower enriched fuel determined a limit of 25 fuel rods loaded into each fuel assembly channel of the package. Using this fuel rod limit and the worse-case parameters, the maximum system reactivity is 0.9094, i.e., below the USL of 0.9340.
The applicant also determined the criticality safety of 10x10 LEU+ UO2 fuel rods contained in a 5-inch stainless steel pipe. The model for containing fuel rods within a stainless-steel pipe uses a triangular lattice for rod configuration and doesnt credit the stainless steel for criticality safety.
A fuel rod pitch sensitivity study was performed by the applicant with the results showing keff values higher than the USL of 0.94254; therefore, the applicant applies a limit on the number of rods. For 10x10 LEU+ fuel, this limit is 30 rods.
To determine the packages sensitivity to hydrogenous packing material on criticality, the applicant completed a study on polyethylene packing material location and density. Due to increased interaction between packages, the applicant demonstrated that the outer container 5
being voided has the highest reactivity. Therefore, limits on packing material mass outside of the stainless-steel pipe are not warranted. However, within the pipe, due to the highest reactivity being full density water and polyethylene having a hydrogen atom density slightly less than that of full density water, the applicant determined that a mass limit needs to be applied. For 30 loose 10x10 UO2 LEU+ fuel rods, the applicant found this limit to be 32.1 kg of polyethylene with a reference density of 0.80 g/cm3 within the stainless-steel pipe, and unlimited otherwise. With these conditions applied, the applicant determined that the most limiting keff + 2 value is 0.9143, i.e., below the USL of 0.94254.
The applicant used the SCALE 6.1 code system, with the KENO VI three-dimensional Monte Carlo code and the continuous-energy ENDF/B-VII.0 cross section library, for all criticality analyses. The applicants previously approved benchmarking analysis for this code and cross-section library determined a USL of 0.9340 for LEU fuel assemblies and loose rods (<5.0 wt%).
The applicant kept the same USL calculation for the LEU+ fuel due to the critical benchmark experiments having enrichments that bound the 8.0 wt% fuel contents. The requested enrichment is within the range of applicability of the previous benchmarking analysis. The staff performed an independent analysis of the USL calculations and bias determination and found the applicants analysis acceptable. The staff finds the USL calculation acceptable for the criticality safety of the LEU+ fuel rods.
Previously, the Criticality Safety Index (CSI) for this package was 1.0 for fuel assemblies enriched to 5.0 wt%. For the RAJ-II package containing fuel assemblies enriched up to 8.0 wt%,
the criticality analysis demonstrates safety in 21x3x24 arrays (1,512 packages; N=302) for packages under NCT and 8x8 arrays (64 packages; N=32) for damaged packages under HAC, giving a calculated CSI of 1.6 for any RAJ-II package shipped with an enrichment greater than 5.0 wt%.
The staff performed confirmatory analyses using the SCALE 6.2.4 Monte Carlo radiation transport code, with the CSAS6 criticality sequence and the continuous-energy ENDF/B-VII.1 neutron cross section library. Using modeling assumptions similar to the applicants, the staffs independent evaluation resulted in keff values that were similar to, or bounded by, the applicants results.
The staff finds that the applicant has identified the most reactive configuration of the RAJ-II package with the new LEU+ fuel assembly and loose rod contents within the GNF 10x10 assembly, and that the calculated keff results are conservative.
Therefore, the staff finds with reasonable assurance that the package, with the requested contents, will meet the criticality safety requirements of 10 CFR Part 71.
CONDITIONS
The following changes were made to the Conditions of the certificate:
Item No. 3.b reflects the latest revision (revision No. 11) of the application, as supplemented.
Condition No. 5(a)(2) was modified to add or aluminum to the sentence pertaining to the shock absorbers.
Condition No. 5(a)(3) was modified to include new revisions of licensing drawings and also add the Outer Container Drawing 105E3738 Sheets 2 and 3, revision 10.
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Condition No. 5(b)(1) was modified to replace the wording of reprocessed uranium with slightly contaminated uranium with trace quantities limits, and state that 238U, not listed in table 2, is assumed to be the remainder of the uranium concentration. Table 2 has been updated to include the maximum authorized concentrations for 8% wt. U235. The previously numbered table 3 was split two separately numbered tables, one for CSI=1.0 (table 3) and a separate table for CSI=1.6 (table 4). Table 3 title is now "Fuel Assembly Parameters (CSI=1.0)" and table 4 is titled "Fuel Assembly Parameters (CSI=1.6)". Table references in the CoC are updated accordingly. The previously numbered table 4 is now numbered table 5. Table 5 was revised to account for the new values of the polyethylene equivalent masses: the mass for 8x8, 9x9, and 10x10 UO2 is unlimited in all cases for rods less than 5 wt.% 235U. For the BWR 10x10 rods greater than 5 wt.% but less than 8 wt.% 235U, there is a polyethylene mass limit of 32.1 kg when these rods are stored within the stainless-steel pipe. In all other configurations, the polyethylene mass is unlimited (i.e., there is a mass limit in the stainless-steel pipe when the enrichment X is 5 wt% X 8 wt% 235U).
Condition No. 5(b)(2) was modified to include contents > 5.0 wt% and 8.0 wt% 235U
Condition No. 5(c) was updated to include the revised CSI of 1.6 for contents > 5.0 wt% and 8.0 wt% 235U in table 4, as well as for the fuel rods as stated in table 5.
Condition No. 11 authorizes the previous revision of the certificate up until the expiration date of the certificate rev. 12, i.e., January 31, 2024.
Condition No. 12 was updated to show the new expiration date of the certificate of January 31, 2029.
The reference section of the certificate was modified to indicate revision No. 11 of the application dated September 2022, as supplemented on June 12, 2023.
CONCLUSION
Based on the statements and representations in the application, the staff finds that these changes do not affect the ability of the package to meet the requirements of 10 CFR Part 71.
Issued with CoC No. 9309, Revision No. 13.