ML24156A104
| ML24156A104 | |
| Person / Time | |
|---|---|
| Site: | 07109372 |
| Issue date: | 06/06/2024 |
| From: | Storage and Transportation Licensing Branch |
| To: | Framatome |
| Shared Package | |
| ML24156A102 | List: |
| References | |
| Download: ML24156A104 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001
SAFETY EVALUATION REPORT Docket No. 71-9372 Model No. TN-B1 Certificate of Compliance No. 9372 Revision No. 5
SUMMARY
By letter dated January 31, 2023 ( Agencywide Documents Access and Management System
[ADAMS] Accession No: ML23031A213), Framatome Inc. (the applicant), requested an amendment to Certificate of Compliance (CoC) No. 9372, for the Model No. TN-B1 transportation package to support Framatomes advancement to enrichments greater than 5 weight percent (wt.%) 235U.
On March 22, 2024, the applicant responded to staffs request for additional information dated November 1, 2023 (ML23321A179). The applicant provided updated calculations including the additional benchmarks establishing the upper safety limit (USL) as requested by the staff. The calculation determines the USL conservatively moves from 0.9318 to 0.9325. As a result, there are no changes to the gadolinia requirements or criticality safety index values previously presented in the application.
The certificate has also been renewed in accordance with the timely request for renewal letter dated December 14, 2023.
The NRC staff reviewed the applicants request and found that the package meets the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.
EVALUATION
The application, i n its revision No. 11, includes the following changes:
(i) identification of two enrichment ranges (<5% and <8% wt.% 235U) for the package contents and discussion of the new 17x17 Type 3 pressurized-water reactor (PWR) fuel rod criteria, (ii) revision of drawings to include varying size s and quantities of the vibro-isolator rubbers for an enhanced vibration performance, (iii) updated leak rate information corresponding to the higher enrichment contents.
The fuel assemblies in this packaging, with two enrichment ranges up to a maximum of 8.0 wt.%
235U, include both Type A and Type B materials. If the contents consist of Type B quantities of material, the main nuclides are shown in the Table below.
Enclosure 2
Type B Quantity of Radioactive Material (Type A and B) for 8.0 wt.% 235U
Specific Total Activity, Total Activity, Isotope Maximum content4 Maximum Activity5, TBq/g TBq Ci mass, g U-232 1.10 10-7g/gU 1.10 10-7 0.83 4.53 10-2 1.20 U-234 7.65 10-3g/gU 7.65 10-3 2.3 10-4 8.73 10-1 23.5 U-235 8.00 10-2 g/gU 8.00 10-2 8.00 10-8 3.17 10-3 8.73 10-2 U-236 2.50 10-2 g/gU 2.50 10-2 2.40 10-6 2.98 10-2 8.06 10-1 U-238 8.87 10-1 g/gU 8.87 10-1 1.20 10-8 5.28 10-3 1.50 10-1 NP-237 1.66 10-6 g/gU 1.66 10-6 2.60 10-5 2.14 10-5 5.85 10-4 PU-238 6.20 10-11 g/gU 6.20 10-11 6.30 10-1 1.94 10-5 5.23 10-4 PU-239 3.04 10-9 g/gU 3.04 10-9 2.30 10-3 3.47 10-6 9.35 10-5 PU-240 3.04 10-9 g/gU 3.04 10-9 8.40 10-3 1.27 10-5 3.47 10-4 Gamma 5.18 10+5 N/A N/A 2.57 10-2 6.94 10-1 Emitters MeV-Bq/kgU Total 2.82 10-1 26.5
Drawing No. 105E3739 Rev. 4 was replaced with Drawing No. FS1 -0042699 Rev.1 to allow multiple rubber options with the goal of optimizing vibration reduction characteristics of the rubber material during transport. Drawing No. FS1 -0042700 Rev. 2 was revised to include a new Note 4 regarding the optional vibro-isolator size and quantity to optimize vibration absorption options.
The full list of licensing drawings for this package is below:
Drawing number Number of Revision # Name Sheets 105E3737 1 6 Outer/Inner Container Assembly Licensing Drawings FS1-0042698 4 1 TN-B1 Outer Container Main Body Assembly Licensing Drawings FS1-0042699 2 1 Outer Container Fixture Assembly Licensing Drawings FS1-0042700 1 2 TN-B1 Outer Container Fixture Assembly Installation Licensing Drawings 105E3741 1 1 Outer Container Shock Absorber Assembly Licensing Drawings 105E3742 1 3 Outer Container Bolster Assembly Licensing Drawings
FS1-0042703 1 1 TN-B1 Outer Container Lid Assembly Licensing Drawings 02-9162717 1 1 Outer Container Marking Licensing Drawings
Inner Container Drawings
Drawing number Number of oevi me ps FS1-0042705 4 1 TN-B1 Inner Container Main Body Assembly Licensing Drawings 105E3746 1 1 Inner Container Parts Assembly Licensing Drawings FS1-0042707 2 1 TN-B1 Inner Container Lid Assembly Licensing Drawings FS1-0042708 1 1 TN-B1 Inner Container End Lid Assembly Licensing Drawings 02-9162722 1 1 Inner Container Marking Licensing Drawings
Contents Container Drawings
Drawing number Number of oevi me ps 105E3773 1 1 Protective Case 0028B98 1 1 Shipping Container Loose Fuel Rods
The applicant added clarification wording to Section 6.0 and 6.1 of the application, along with adding new gadolinium oxide (Gd2O3) requirements for the added enrichment percentages between 5.0 and 8.0 weight percent (wt.%) uranium-235 (235U) in Table 6-1. The applicant also added the new technical specifications of the PWR 17x17 Type 3 fuel assembly type and the loose rod limits packed in a 5-inch SS pipe or protective case within Table 6-2.
Table 6-3, the Critical Evaluation Summary, was amended by the applicant to include the normal conditions of transport (NCT) and hypothetical accident conditions (HAC) k-eff values, along with the number of packages used within the evaluation for the ATRIUM 11 assemblies, AT R I U M 11 rods, ATRIUM 10 rods, the PWR 17x17, type 3 rods, and the criticality safety index (CSI) values for the 11x11 assemblies, the 11x11 fuel rods contained in a 5-inch steel pipe, and the PWR 17x17 type 3 fuel rods all containing 8.0 wt.% 235U enrichment.
The applicant revised the application to include a new Appendix 6.13, which evaluates the criticality safety for the package containing Atrium 11 fuel assemblies, ATRIUM 11 fuel rods, and individual PWR 17x17, type 3 fuel rods all with a maximum enrichment of 8.0 wt.% 235U.
Also, the applicant included additional limits for transport of Atrium 10XM fuel rods with a maximum enrichment of 5.0 wt. % 235U. The models geometry was unchanged from previous versions of the application.
Previously the Gd 2O3 wt.% minimum was 2.0 wt.%, and the scoping studies performed by the applicant supported this minimum Gd2O3 content up until a fuel enrichment of 5.0 wt.% 235U.
With the amendment request to increase fuel enrichment to 8.0 wt.% 235U, the Gd2O3 wt %
minimum was increased to 4.0 wt.% to maintain a maximum CSI of 3.3 for ATRIUM 11 assemblies.
A study was conducted by the applicant to determine the Gd Rod l oading pattern which has the greatest reactivity, the results of which are shown in Figure 6-87 of the application. The applicants loading study used the design requirements of the orientation such that the Gd rods must be placed symmetrically with respect to the assembly major diagonal and the Gd rods could not be placed on the periphery of the assembly. For enrichments above 5 wt.% 235U, each face of the enriched lattices must contain a minimum of one equivalent Gd rod in the second and tenth row/column locations.
Using the Gd rod pattern, the most reactive fuel assembly orientation within the TN-B1 was determined by the applicant as shown in Figure 6-88 of the application. Studies for fuel channel position, polyethylene pad loss, preferential flooding of the package, and flooding between packages were performed by the applicant. A maximum of 10.2 kg of polyethylene per assembly can be contained within the TN-B1. When modeling, SCALE limitations dictate that the polyethylene be homogenized either with the clad or the water to perform resonance self-shielded multi-group cross-section calculations. The applicant determined that for the ATRIUM 11 fuel assemblies in an HAC array, the most reactive model is the polyethylene layered on clad and homogenized with the clad for determining cross sections.
Material compositions within the model were determined by the applicant using a previously approved process based on 98.2% of the UO2 maximum theoretical density (MTD). For fuel rods containing Gd, the overall density of the fuel was calculated as a linear combination of the MTD UO2 and Gd2O3 applying the 98.2% factor to UO2. The applicant credits 75% of the Gd wt.%, therefore the models contain 1.5 wt.% or 3 wt.% of Gd2O3. The polyethylene layer thickness for the ATRIUM 11 fuel assembly model was determined by the applicant using the outer radius of the fuel rod clad, polyethylene density, and total mass of polyethylene. The applicant modeled the homogenized clad and polyethylene along with the homogenized water and polyethylene by determining volume fractions of each component and weighting the homogenized mixture by the volume fractions. A SCALE user notice [REF] identified an under-prediction of k-eff values in systems containing h1-poly. The applicant adjusted their k-eff values by adding a poly bias to consider the possible underprediction.
There are three transport cases for this package:
(i) containing ATRIUM 11 fuel assemblies, (ii) containing loose fuel rods, and (iii) containing loose fuel rods within a stainless-steel pipe.
For an ATRIUM 11 assembly single package, the applicant evaluated the package under NCT with water in-leakage between the TN-B1 inner and outer containers. The assembly contained
the highest enrichment, 8 wt.% 235U, and the lowest wt.% and number of Gd rods, 2 wt.% Gd2O3 and 13 rods. The most reactive orientation of the fuel channel and the location of the polyethylene was used. The resulting k-eff under NCT was 0.7514, under the applicants calculated Upper Subcritical Limit (USL) of 0.9325.
For an ATRIUM 11 assembly single package, the applicant evaluated the package under HAC with full density water between the inner and outer container, the cushioning pads replaced with water, and the fuel channel location chosen for maximum reactivity. The assembly contained the highest enrichment, 8 wt.% 235U, and the lowest wt. % and number of Gd rods, 2 wt.% Gd2O3 and 13 rods. The resulting k-eff under HAC was 0.85127, below the USL of 0.9325.
The staff reviewed the applicants single package evaluations and finds that the applicant has demonstrated that a single package with water in-leakage is subcritical per 10 CFR 71.55(b),
and that a single package is subcritical under NCT and HAC per 10 CFR 71.55(d) and (e),
respectively.
For package arrays under HAC, the applicant modeled a finite 4x1x8 array of packages containing ATRIUM 11 fuel assemblies. The applicant found and used the most reactive water density and the most reactive preferential flooding pattern between each package. The preferential flooding pattern includes the internal compartment flooded with full density water and the volume between the inner and outer compartments empty. Instead of adjusting array size until the USL is approached, the applicant had a CSI goal of 3.2 and found the Gd rod limit for each enrichment between 5 and 8 wt.%. Each k-eff value includes bias and uncertainties.
The results are summarized in Table 1 below.
Table 1. Enrichment and Minimum Gd rods for ATRIUM 11 HAC Array wt.% 235U Minimum Gd Rods Max k-e ff USL 6.3 13 @ 2 wt.% 0.9300 0.9325 6.7 15 @ 2 wt.% 0.9301 0.9325 6.6 13 @ 4 wt.% 0.9304 0.9325 7 15 @ 4 wt.% 0.9289 0.9325 7.6 17 @ 4 wt.% 0.9307 0.9325 8 19 @ 4 wt.% 0.9255 0.9325
The k-eff value for each combination of enrichment and the minimum number and absorber concentration of Gd rods is below the USL of 0.9325.
For package arrays under NCT, the applicant modeled a finite 9x1x9 array of packages containing ATRIUM 11 fuel assemblies. The model still used the same preferential flooding pattern within the packages, but differed by keeping the polyethylene foam pads intact, maintaining a larger outer container height, and reducing the fuel rod spacing to the undamaged value. The results are summarized in Table 2 below.
Table 2. Enrichment and Minimum Gd rods for ATRIUM 11 NCT Array Wt.% 235U Minimum Gd Rods Max k-e ff USL 5.8 13 @ 2 wt.% 0.9295 0.9325 6.1 13 @ 4 wt.% 0.9311 0.9325 6.5 15 @ 4 wt.% 0.9308 0.9325 7 17 @ 4 wt.% 0.9299 0.9325 7.5 19 @ 4 wt.% 0.9292 0.9325 8 21 @ 4 wt.% 0.9252 0.9325
The k-eff value for each combination of enrichment and the minimum number and absorber concentration of Gd rods is below the USL of 0.9325.
The second case the applicant modeled was for loose fuel rods in the TN-B1 package, including ATRIUM 11, ATRIUM 10XM, and PWR 17x17 type 3 fuel rods. Based on sensitivity studies done before, the most reactive configuration contained:
- TN-B1 central storage volumes flooded with full density water
- Volume between outer and inner TN-B1 containers empty
- Full thickness of polyethylene pads
- 25 full-length fuel rods in a 5x5 square-pitched array
- Fuel rod gap and clad modeled as full density water
- Evaluated with and without polyethylene sleeve to determine h-poly bias
- No credit for Gd2O3
The single package results are summarized in Table 3 below.
Table 3. Single Package Results for Loose Rods Fuel Rod Design HAC Max Keff NCT Max Keff ATRIUM 11 0.6832 0.6647 ATRIUM 10XM 0.6328 0.6181 PWR 17x17 0.6838 0.6656
The k-eff values for each single package fuel rod design under HAC and NCT are well below the USL of0.9325.
The staff reviewed the applicants single package evaluations for loose rods and finds that the applicant has demonstrated that a single package with water in-leakage is subcritical per 10 CFR 71.55(b), and that a single package is subcritical under normal conditions of transport and hypothetical accident conditions per 10 CFR 71.55(d) and (e), respectively.
Similarly to the ATRIUM 11 evaluation, the NCT array for loose fuel rods differed from the HAC array by increasing the height of the outer container according to the results of the HAC structural tests. For package arrays under NCT, the applicant modeled a finite 16x1x16 array, and for package arrays under HAC, the applicant modeled a finite 10x1x10 array. The maximum k-eff for each fuel rod type is summarized below.
Table 4. K-eff Values for Loose Rods NCT Fuel Rod Design Max Keff i ATRIUM 11 0.8904 0.9325 ATRIUM 10XM 0.8255 0.9325 PWR 17x17 0.8937 0.9325
Table 5. K-eff Values for Loose Rods HAC Fuel Rod Design Max Keff i ATRIUM 11 0.8652 0.9325 ATRIUM 10XM 0.8024 0.9325 PWR 17x17 0.8681 0.9325
For the fuel rod designs above, the maximum k-eff values are below the USL of each system.
The applicant determined the CSI for each of the above contents of the TN-B1 package according to the requirements for package arrays in 10 CFR 71.59. For NCT, the applicant showed that an array of 256 packages is subcritical. For HAC, the applicant showed that an array of 100 packages is subcritical. Using the most limiting case, the applicant determined the CSI to be 1.0.
The staff finds that the applicant has appropriately determined the package CSI in accordance with the requirements of 10 CFR 71.59.
The third case the applicant modeled was for ATRIUM 11 rods ( 8.0 wt.% 235U), ATRIUM 10XM rods ( 5.0 wt.% 235U), and PWR 17x17, type 3 rods ( 8.0 wt. % 235U) contained in either a 5-inch schedule 40 stainless steel pipe or a protective carrier. Both the pipe and the protective carrier have been approved in previous amendments. The applicant conducted sensitivity studies to determine the most reactive configuration, which includes:
- Optimum triangular pitch for the fuel rods within the pipe.
- Optimum water density for water that is inside the inner package and outside the pipe.
- Optimal model for polyethylene sleeves.
The applicants HAC model determined the optimal peripheral pad thickness and determined the optimal pipe locations within the internal transport volume. The most reactive pin pitch was determined to be quite large, and because the protective cases cross-sectional area is much smaller than the stainless-steel pipes cross-sectional area, the transport of loose rods is bounded by the stainless-steel pipe geometry. Similar to the loose fuel rods case, the NCT model differed by increasing the height of the outer container and keeping full thickness of the peripheral cushioning pads.
The results of the applicants HAC and NCT single package evaluations are summarized below.
The polyethylene error was added when applicable, along with the calculations biases and uncertainties.
Table 6. Loose Rods in Pipe, HAC and NCT Single Package Results Fuel Rod Design HAC Single Package Keff NCT Single Package Keff ATRIUM 11 0.7630 0.7014 ATRIUM 10XM 0.7114 0.6531 PWR 17x17 0.7655 0.7043
All maximum k-eff values are well below the USL of 0. 9325.The staff reviewed the applicants single package evaluations for loose rods and finds that the applicant has demonstrated that a single package with water in-leakage is subcritical per 10 CFR 71.55(b), and that a single package is subcritical under normal conditions of transport and hypothetical accident conditions per 10 CFR 71.55(d) and (e), respectively.
To determine CSI values, the applicant modeled arrays of the packages under HAC and NCT.
The models and their subsequent k-eff values and CSI values are summarized below.
Table 7. NCT and HAC CSI Calculations for Loose Rods in Pipes Fuel Rod HAC Keff CSIHACNCT Keff CSINCT Design Array Array ATRIUM 9x1x9 0.9240 1.3 10x1x11 0.9285 2.3 11 ATRIUM 10x1x10 0.8690 1.0 11x2x12 0.8978 1.0 10XM PWR 9x1x9 0.9266 1.3 10x1x10 0.9241 2.5 17x17
Taking the most limiting CSI of each case, the applicant determined that for the TN-B1 loaded with loose fuel rods in a pipe, the CSIs for ATRIUM 11, ATRIUM 10XM, and the PWR 17x17 fuel rods are 2.3, 1.0, and 2.5 respectively. The staff finds that the applicant has appropriately determined the package CSI in accordance with the requirements of 10 CFR 71.59.
The staff reviewed the configurations modeled by the applicant for the single package and array analyses. The staff finds with reasonable assurance that the applicant has identified the most reactive credible condition of the single package and arrays of packages, consistent with the condition of the package under NCT and HAC, and the chemical and physical form of the fissile and moderating contents.
The applicant selected applicable benchmark experiments to validate the ATRIUM 11 contents enriched up to 8.0 wt.% and loose PWR rod contents enriched up to 8.0 wt.% using sensitivity/uncertainty analysis methods (S/U).
The applicant used the TSUNAMI-3D sequence included in the SCALE 6.2.4 code package to calculate the sensitivity of the k-eff value from the TN-B1 package bounding array case (known as the application model) for both the ATRIUM 11 and loose PWR rods to variations of the nuclear data used in the k-eff calculation. The TSUNAMI-3D sequence generates sensitivity
data files (SDFs) containing sensitivities of k-eff to variations in cross section data for each reaction type. The applicant then used TSUNAMI-IP sequence to compare the application SDF against potential critical benchmark experiment SDFs. The TSUNAMI-IP sequence generates correlation coefficients (ck values) that indicate similarity between the application model and the critical benchmark experiment. The applicant only used this S/U method to select applicable benchmarks for validation.
When using TSUNAMI-3D it is important to ensure that the model is divided into an appropriate grid size to determine the sensitivities, and determining the right grid size can be difficult. The need for finer grids for fissile material regions must be balanced against the large computer memory requirement and longer computation time for a smaller grid size.
The applicant calculated sensitivities for both 235U and 1H using direct perturbations and compared them to the sensitives generated using TSUNAMI-3D.
For the package containing the ATRIUM 11 assembly, sensitivity coefficients using direct perturbation were 0.201 (235U) and 0.054 (1H) and using TSUNAMI-3D were 0.182 (235U) and 0.062 (1H).
For the package containing loose PWR rods, sensitivity coefficients using direct perturbation were 0.220 (235U) and 0.258 (1H) and using TSUNAMI-3D were 0.209 (235U) and 0.207 (1H).
The staff reviewed the applicants use of S/U methods to ensure they were applied appropriately.
The experiments selected by the applicant for validation can be found in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). For the package containing the ATRIUM 11 assembly, the applicant selected a total of 254 experiments with a minimum ck value of 0.80, of which 27 had ck values > 0.90.
For the package containing loose PWR rods, the applicant selected a total of 241 experiments with a minimum ck value of 0.80 of which 96 had ck values > 0.90. Table 6-155 of Appendix 6.13.10.1 of the application includes comparisons between the critical benchmark experiments and the TN-B1 package with the ATRIUM 11 assemblies, demonstrating the applicability of the included benchmarks. Table 6-156 of Appendix 6.13.10.1 of the application includes comparisons between the critical benchmarks experiments and the TN-B1 package with loose PWR rods demonstrating the applicability of the included benchmarks. Many methods used to calculate bias and bias uncertainty (used to determine the USL) rely on the assumption that the population of critical experiments constitutes a normal distribution. Since the data set does not follow a normal distribution, the applicant applied a non-parametric technique that uses an analysis of ranks to determine the USL.
For the ATRIUM 11 assembly sample population size, the rank index for a one-sided distribution-free tolerance limit with 95% confidence that 95% of the population is covered is 7, meaning the seventh lowest calculated k-eff value, which for the population of experiments selected by the applicant is 0.98644. Including an administrative margin of 0.05 yields a USL of 0.9325.
For the loose PWR rod sample population size, the rank index for a one-sided distribution-free tolerance limit with 95% confidence that 95% of the population is
covered is also 7, The seventh lowest calculated k-eff value is also 0.98644, obtaining the same USL of 0. 9325.
The staff finds that the applicant determined an appropriate USL using S/U methods. Staff S/U analysis and USL determination confirms that the applicants USL is conservative.
The staff performed confirmatory calculations using the SCALE 6.2.4 Monte Carlo radiation transport code, with the CSAS6 criticality sequence and the continuous energy ENDF/B-VII neutron cross section library. The staffs confirmatory analyses focused on the most reactive configuration of single packages and arrays of packages. Using modeling assumptions similar to the applicants, the staffs evaluation resulted in keff values that were similar to, or bounded by, the applicants results.
The staff contracted with Oak Ridge National Laboratory (ORNL) to perform confirmatory benchmarking calculations for validation purposes. ORNL used TSUNAMI-3D to generate SDFs from both the ICSBEP Handbook and from the ORNL SCALE Verified, Archived Library of Inputs and Data (VALID), and TSUNAMI-IP to generate ck values between the application model and critical experiments. From the VALID library, ORNL found 69 experiments with ck values >
0.8 for the ATRIUM 11 assembly model, and from ICSBEP Handbook, 187 experiments for the ATRIUM 11 assembly and 154 for loose PWR rods with ck values > 0.8. Neither experiment set had a normalized distribution and a non-parametric technique as described by the applicant was used by ORNL to determine the USL.
For the ATRIUM 11 assembly, ORNL determined USLs of 0.94340 (VALID) and 0.93242 (ICSBEP), both including the administrative margin of 0.05.
For the loose PWR rods, ORNL determined USLs of 0.94340 (VALID) and 0.93460 (ICSBEP), both including the administrative margin of 0.05.
The USLs generated by ORNL are similar to or greater than the applicants determined USLs of 0.9325 (ATRIUM 11 and loose PWR rods) which demonstrates that the applicants USL is appropriate.
Based on the discussion above, the staff found the changes to the CoC would not affect the ability of the TN-B1 package to meet the criticality safety requirements of 10 CFR Part 71.
CONDITIONS
The following changes have been made to the CoC:
Item No. 3.b. has been updated to include application FS1-0014159, Revision 11, as supplemented on March 22, 2024.
Condition No. 5(a)(2) was revised to include the maximum enrichment of 8 wt.%.
Condition No. 5(a)(3) was revised due to drawings being replaced. A new drawing FS1-0042699-1.0 allows multiple rubber options and drawing FS1- 0042700- 2.0 was revised.
Condition No. 5(b(1)(ii) was added to include contents > 5.0 to 8.0 wt. % U-235. A new Table 3 shows the maximum concentrations of authorized contamination material for > 5.0 to 8.0 weight percent U-235.
Conditions No. 5(b)(1)(v) and (viii) were added for 11x11 fuel assemblies and fuel rods respectively. Tables 6 and 8 are new and give the fuel assemblies and fuel rods parameters, respectively, for > 5.0 to 8.0 weight percent U-235.
Condition No. 5(b)(2) has been revised.
Condition No. 5(c) has been entirely revised.
Condition No. 11 has been revised to extend the validity of the certificate to June 30, 2029.
The references section has been updated to include the March 22, 2024, supplement to the application.
CONCLUSION
Based on the statements contained in the application, and the conditions listed above, the staff concludes that the changes indicated do not affect the ability of the package to meet the requirements of 10 CFR Part 71.
Issued with Certificate of Compliance No. 9372, Revision No. 5.