ML18152A983
| ML18152A983 | |
| Person / Time | |
|---|---|
| Site: | 07106386 |
| Issue date: | 05/31/2018 |
| From: | John Mckirgan Spent Fuel Licensing Branch |
| To: | Pratt R US Dept of the Navy |
| White B | |
| Shared Package | |
| ML18152A982 | List: |
| References | |
| EPID L-2017-LLA-0062, USA/6386/B(U)F-96 | |
| Download: ML18152A983 (7) | |
Text
May 31, 2018 Mr. Roy G. Pratt NAVSEA 08G 1240 Isaac Hull Ave SE Stop 8021 Washington Navy Yard, DC 20376-8021
SUBJECT:
REVISION NO. 22 OF CERTIFICATE OF COMPLIANCE NO. 6386 FOR THE MODEL NO. 235R001 TRANSPORTATION PACKAGE
Dear Mr. Pratt:
As requested by Naval Reactors application dated June 9, 2017, enclosed is Certificate of Compliance (CoC) No. 6386, Revision No. 22, for the Model No. 235R001 transportation package. Changes made to the enclosed certificate are indicated by vertical lines in the margin.
The U.S. Nuclear Regulatory Commission staffs safety evaluation report is also enclosed.
If you have any questions regarding this certificate, please contact Bernard White of my staff at 301-415-6577.
Sincerely,
/RA/
John McKirgan, Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-6386 EPID No. L-2017-LLA-0062
Enclosures:
- 1. CoC No. 6386, Rev. No. 22
- 2. Safety Evaluation Report
R. Pratt REVISION NO. 22 OF CERTIFICATE OF COMPLIANCE NO. 6386 FOR THE MODEL NO.
235R001 TRANSPORTATION PACKAGE, DOCUMENT DATE: May 31, 2018 DISTRIBUTION:
G:\SFST\PART 71 CASEWORK\Model 235R001 Docket 71-6386\6386 r22.LTR&SER.doc G:\SFST\PART 71 CASEWORK\Model 235R001 Docket 71-6386\6386.r22.doc ADAMS Package No.: ML18152A982 Letter & SER.: ML18152A983 CoC: ML18152A984 OFC: DSFM DSFM DSFM DSFM NAME: BWhite SFigueroa SEverard JSmith Via email via email via email DATE: 4/23/18 5/7/18 4/24/18 5/4/18 OFC: DSFM DSFM DSFM NAME: TTate YDiaz-Sanagria JMcKirgan via email via email DATE: 5/8/18 5/15/18 5/31/18 OFFICIAL RECORD COPY
SAFETY EVALUATION REPORT Docket No. 71-6386 Model No. 235R001 Certificate of Compliance No. 6386 Revision No. 22 EVALUATION By application dated June 9, 2017, the U.S. Department of Energy, Naval Reactors, requested and amendment to Certificate of Compliance (CoC) No. 6386, for the Model No. 235R001 transportation package. The applicant requested deletion of the S3G fuel refueling cells as authorized contents and changes to various criticality safety indexes (CSIs). The NRC has reviewed the amendment and finds that the changes do not affect the ability of the package to meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.
The NRC staff performed its review of the amendment to the 235R001 certificate of compliance utilizing the guidance provided in NUREG-1609, Standard Review Plan for Transportation Packages for Radioactive Material.
Structural Evaluation The Naval Reactors Program uses the 235R001 shipping containers to ship A1G fresh fuel cells to servicing facilities. During the development of the A1G configured M-290 spent fuel transportation package safety analysis report (A1G SAR), the applicant determined that the A1G lead screw and tie-rod were susceptible to brittle fracture failure in the M-290 shipping container as a result of the hypothetical accident conditions in 10 CFR 71.73. In addition, the applicant stated that this failure mode had not been considered for the A1G fresh fuel assembly in the 235R001 shipping container SAR (A1G 25-181 SAR). Specifically, failure of the lead screw invalidates the control rod position assumption used in previous criticality analyses.
In addition to the lead screw analysis, the applicant also stated that the A1G fuel cells are planned to transition to a lower enrichment. As a result, the applicant performed an analysis of the lower enriched A1G fuel assembly in the 235R001 shipping container against the requirements of 10 CFR 71.
The scope of the structural safety evaluation is limited to reviewing the applicants analysis of the lead screw and their assumptions for the position of the control rod used in the criticality analysis under both normal conditions of transport and hypothetical accident conditions, as well as evaluating the analysis of the lower enriched A1G fresh fuel transported in the 235R001 shipping container.
Enclosure 2
2.1 Description of Structural Design With the exception of the enrichment of the fuel cells, neither the structural design of either the Naval Reactors 235R001 shipping container, nor the A1G contents were changed via this amendment.
2.2 Normal Conditions of Transport In Appendix (1) of Enclosure (12) of the A1G 25-181 SAR, the applicant compared the higher and lower enriched A1G fuel cells with respect to materials, dimensions and weight to determine any differences that would warrant a new analysis. The applicant compared the drawings of both fuel cell types and determined that the specified materials, specified dimensions, and weights were the same for both fuel types (with the exception of the enrichment of the fuel material). The applicant stated that the weights of the fuel clusters often have a weight variability due to construction tolerances, which includes different fuel loadings. Although the offset between the fuel loadings may result in a slightly different weight for the lower enriched fuel, it is well within the specified tolerance, which are considered in all models and analyses.
As a result of their analysis, the applicant determined that the structural calculations for the package of the higher enriched fuel bound those of the lower enriched fuel and that shipment of the lower enriched fuel was covered by the existing structural analysis.
The staff reviewed the applicants analysis and determined that all structural analyses conducted on the higher enriched fuel, to include finite element models, bound all applicable analyses for the lower enriched fuel and that the structural assumptions for normal and accident condition package configurations for criticality evaluations remain unchanged. Because of this, the staff finds that the 235R001 transportation package, loaded with the lower enriched A1G fuel cells, continues to satisfy the applicable structural requirements of 10 CFR 71.
In Appendix (2) of Enclosure (12) to the A1G 25-181 SAR, the applicant determined the load capacity of the lead screw with respect to brittle fracture and compared this value to the loads encountered during normal conditions of transport. To do this, the applicant first determined the stress required to produce a brittle fracture in the lead screw material. This level of stress occurs when the brittle fracture stress intensity factor equals the fracture toughness of the material. The applicant then used this stress value to determine the inertial load required to achieve this level of stress in the material. The applicant then compared this load capacity with the resultant inertial loads produced by the various tests proscribed for normal conditions of transport in 10 CFR 71.71. The applicant stated that the inertial load required to produce a brittle fracture in the lead screw material was greater than two times the load produced by normal conditions of transport (vibration, shock, and the 3-foot flat bottom drop tests).
The applicant also noted that four decades of acceptable 235R001 shipments of A1G fresh fuel cells provides supplementary evidence that the A1G cells will not be affected by normal conditions of transport loads.
The staff reviewed the applicants calculations for determining the inertial load required to produce a brittle fracture in the lead screw material. Because the load required to produce a brittle fracture is more than twice that produced by the proscribed normal conditions of transport tests, the staff finds that there is no substantial reduction in the effectiveness of the packaging that would prevent it from satisfying the requirements of 10 CFR 71.55(d)(2) for a fissile material package.
2.3 Hypothetical Accident Conditions Because the applicant conceded lead screw failure during the hypothetical accident conditions tests in 10 CFR 71.73, it did not provide further structural calculations on the A1G fuel assembly and assumed complete control rod withdrawal for the subsequent criticality evaluation.
Based on the applicants statement in the SAR, the staff finds that the geometric form of the package contents is substantially altered under the hypothetical accident conditions tests in 10 CFR 71.73, in that failure of the lead screw due to brittle fracture could cause complete withdrawal of the control rod which will subsequently increase the reactivity of the package.
Further evaluation of the package content with respect to satisfying the subcriticality requirements of 10 CFR 71.55(e) for a fissile material can be found in section 6 of this SER.
2.2 Findings
Based on a review of the statements and representations in the application, the staff concludes that the structural design has been adequately described and evaluated and that the package has adequate structural integrity to continue to meet the requirements of 10 CFR Part 71.
Criticality Evaluation 6.0 Criticality Safety Evaluation The 235R001 shipping container is a transportation package that is designed to ship fresh A1G fuel cells. In addition, the applicant also requests authorization for transport of the A1G fuel cells at a lower enrichment than previously approved.
During the NRC review of the M-290 transportation package, the applicant identified that the A1G lead screw was susceptible to brittle fracture failure in the M-290 transportation package under hypothetical accident conditions, and recognized that this failure mode was not analyzed for criticality safety of the 235R001 shipping container containing the A1G fuel, which could potentially affect the control rod position and, therefore, the reactivity of the contents within the package. The scope of the criticality safety review is limited to the applicants revised criticality safety analysis of the package resulting in altered control rod position due to failure of the lead screw, and evaluation of the proposed lower enrichment of the A1G fresh fuel to be transported in the 235R001 shipping container.
6.1 Description of Criticality Design The criticality design of the 235R001 shipping container remains unchanged from the previously approved analysis. However, the identified lead screw failure during hypothetical accident conditions results in an unanalyzed condition for the potential control rod position in the A1G fuel cells. The applicant addressed this unanalyzed condition by assuming the controls rods are removed from the modeled fuel cells, taking no credit for neutron absorption effects of the control rods.
6.2 Fissile Material Contents The Naval Reactors program is transitioning enrichment of the A1G fuel cells to a lower enrichment than was previously approved. This nominal decrease in enrichment results in a
reduction of the fissile mass contents in the A1G fuel cell, and consequently, a decrease in the neutron multiplication factor in comparison to the original fuel design.
6.3 General Considerations for Criticality Evaluations The applicants analysis is based on the drawings of the A1G fresh fuel package to represent an accurate geometry model, with special importance given to aspects of the model that are important to criticality safety. Material data used in the evaluation include the American Society of Mechanical Engineers, ASTM International, and Naval Nuclear Propulsion Program documents. Calculations are performed using a Monte Carlo neutron transport theory-based computer program, and the model accurately represents the features of the fuel cells in complete detail, including fuel, poison, cladding, coolant channels, control rods and channels, and the structural material of the A1G fuel.
The applicant demonstrated the maximum reactivity of the 235R001 shipping package loaded with any A1G fuel cell by evaluating various sensitivities and assumptions for the single package evaluation under flooded conditions, a single package under normal conditions of transport, a single package under hypothetical accident conditions, and packages in arrays under both normal conditions of transport and hypothetical accident conditions. Based on the application, the CSI calculated for a close packed infinite array of loaded 235R001 packages is 100 in accordance with 10 CFR 71.59(a)(1) and 10 CFR 71.59(a)(2).
6.4 Single Package Evaluation For the single package evaluation, staff confirmed that the applicant adhered to the applicable conditions of 10 CFR 71.55 and evaluated various flooded conditions as well as residual water left in the package. In all instances, the resulting calculated keff that was identified by the applicant was found to be less than 0.95, including all biases and uncertainties for both normal conditions of transport and hypothetical accident conditions.
6.5 Evaluation of Package Arrays Under Normal Conditions of Transport For the array of packages under normal conditions of transport, an infinite close packed hexagonal array was utilized to minimize spacing in an infinite array with the applicant modeling the 235R001 shipping container loaded with any type of A1G fuel cell with varying degrees of moderation consistent with the applicable portions of 10 CFR 71.59. The applicant included a demonstration of maximum reactivity utilizing numerous parametric studies for all of the scenarios evaluated, including the normal location of the control rods. In all instances, the calculated keff was found by the applicant to be less than 0.95, including all biases and uncertainties.
6.6 Evaluation of Package Arrays Under Hypothetical Accident Conditions For the array of packages under hypothetical accident conditions, the applicant modeled a single 235R001 shipping container loaded with any type of A1G fuel cell with varying degrees of moderation consistent with the applicable portions of 10 CFR 71.59. The applicant included a demonstration of maximum reactivity utilizing numerous parametric studies for all of the scenarios evaluated. With respect to the potential for lead screw failure resulting in control rod movement, the control rods were assumed to be fully withdrawn from the A1G fuel cells.
Therefore, no control rod portions were modeled, which increases the reactivity due to less neutron absorption and is a conservative assumption. In all instances of the studies, the
resulting calculated keff were found by the applicant to be less than 0.95, including all biases and uncertainties. The removal of the control rods from the fuel cells results in an increase in the CSI from 25 to 100, which limits transport of the loaded 235R001 shipping container to one package per shipment.
6.7 Benchmark Evaluations The applicant provided an extensive benchmark evaluation that compared calculational methods with experimental results to determine appropriate bias and uncertainties. NRC staff reviewed the benchmarking evaluation methodology, discussed the methodology with Naval personnel at previous site visits to Bettis Atomic Power Laboratory, and the staff determined that the methodology employed complies with the requirements of 10 CFR Part 71.
6.8 Evaluation Finding As a result of the structural review indicating that failure of the lead screw could result in the complete withdrawal of the control rods, the NRC staff reviewed the applicants analysis with regards to the lead screw failure evaluation and the resultant increase in reactivity in criticality analysis. NRC staff also reviewed the applicants analysis of a reduction in fuel enrichment to ensure that the applicants analysis adequately bounded the 235R001 package with this lower enriched A1G fuel. Since the resulting keff for the evaluated system under both normal conditions of transport and hypothetical accident conditions were confirmed by the NRC staff review of the applicants analysis to be less than 0.95, staff concludes that the 235R001 shipping container containing any type of A1G fuel cell under the assumptions utilized by the applicant, and under the conditions listed in the certificate of compliance, continues to meet the criticality safety requirements in 10 CFR Part 71.
CONDITIONS The package description in 5.(a)(2) was revised to remove contents that were removed as authorized contents.
Content in Condition No. 5(b)(1)(iv) for the S3G fuel has been deleted, content Condition Nos.
5(b)(1)(v), 5(b)(1)(vi), 5(b)(1)(vii), and 5(b)(1)(viii) were renumbered throughout.
The CSI for Content No. 5.(b)(1)(i) was changed to 100.
The CSI for Content Nos. 5.(b)(1)(iv), and 5.(b)(1)(v) were reserved for future use.
The references section was updated to include this request.
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. 6386, Revision No. 22, on 5/31/18.