ML19169A077
| ML19169A077 | |
| Person / Time | |
|---|---|
| Site: | 07109341 |
| Issue date: | 06/14/2019 |
| From: | John Mckirgan Spent Fuel Licensing Branch |
| To: | Noss P Orano Federal Services |
| Santos N | |
| References | |
| EPID L-2018-LLA-0117 | |
| Download: ML19169A077 (22) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 14, 2019 Mr. Phil Noss Licensing Manager Orano Federal Services LLC 505 S. 336th Street, Suite 400 Federal Way, WA 98003
SUBJECT:
AUTHORIZATION FOR SHIPMENT OF THE MODEL NO. BRR PACKAGE WITH TEST FUEL SEGMENTS
Dear Mr. Noss:
As requested by your letter dated April 19, 2018 [Agencywide Documents Access and Management System (ADAMS) Accession Number (No.) ML18323A311], as supplemented on September 26, 2018 (ADAMS Accession No. ML18275A113), pursuant to Title 10 of the Code of Federal Regulations Part 71, the Certificate of Compliance (CoC) No. 9341 for the Model No.
BEA Research Reactor (BRR) package is amended by letter to allow test spent fuel assemblies, which are not already authorized contents in the CoC, to be shipped. All other conditions of CoC No. 9341, Revision 7 (ADAMS Package Accession No. ML19038A093) shall remain the same. This authorization is valid for one shipment of a BRR package limited by the following conditions:
- 1. The fuel assemblies will be shipped by truck.
- 2. There will be a maximum of one BRR package per shipment on a single truck.
- 3. The fuel assemblies will only contain commercial uranium dioxide (UO2).
- 4. The fuel assemblies will meet Type B material contents.
- 5. The shipment will only include fuel discharged from 1995 to 2004.
- 6. Fuel rods diameter (including cladding) is approximately 0.4 inches (in.).
- 7. The following table includes additional specific limits about the contents:
Parameter Minimum Maximum Units Total fissile mass --- 10.41 grams (g)
(g non-fissile mass)/
Total non-fissile mass 200 ---
(g fissile mass)
Initial enrichment of uranium-235 (235U) 2.9 4.2 weight percent (wt.%)
Mass of 235U --- 3.14 g Mass of plutonium-239 (239Pu) --- 6.30 g Mass of plutonium-241 (241Pu) --- 0.97 g Maximum weight of the payload --- 20 pound (lbs.)
Megawatts-day per Burnup 48 70 kilogram of uranium (MWd/KgU)
Fuel rod length 1 13 in.
Total length of segments 1 144 in.
Fuel mass per rod 1.78 2.23 kg Decay heat generation --- 3 Watts (W)
P. Noss 2 If you have any questions regarding this authorization, please contact me or Norma García Santos at (301) 415-6999.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
/RA/
John McKirgan, Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-9341 EPID L-2018-LLA-0117
Enclosure:
Safety Evaluation Report
3
SUBJECT:
AUTHORIZATION FOR SHIPMENT OF THE MODEL NO. BRR PACKAGE WITH TEST FUEL SEGMENTS, DOCUMET DATE: June 14, 2019 DISTRIBUTION:
DSFM r/f NMSS r/f NDevaser G:\SFST\Garcia\Casework_NGS\Spent Fuel Packages\BRR-NGS\BRR_Rev_5_Letter_Authorization_2018\Letter authorization_and_SER_BRR_Clean.docx ADAMS Accession No. ML This closed EPID L-2018-LLA-0117.
OFFICE: DSFM DSFM DSFM DSFM DSFM NAME N. García S. Figueroa T. Tate for A. D. Tarantino Y. Kim Santos by email Sotomayor by email by email Rivera DATE: 5/30/19 6/7/19 6/6/19 5/31/19 6/11/19 OFFICE: DSFM DSFM DSFM DSFM DSFM NAME C. Bajwa T. Tate M. Rahimi Y. Díaz J. McKirgan by email by email Sanabria by email DATE: 6/6/19 6/6/19 6/13/19 6/11/19 6/14/19 OFFICIAL AGENCY RECORD
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-9341 Model No. BRR Certificate of Compliance No. 9341 Enclosure
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Table of Contents Page
SUMMARY
................................................................................................................................... 1 EVALUATION............................................................................................................................... 1 1.0 GENERAL INFORMATION ............................................................................................... 1 1.1 Packaging ......................................................................................................................... 1 1.2 Contents ............................................................................................................................ 2 2.0 STRUCTURAL EVALUATION .......................................................................................... 3 2.1 Description of Structural Design .................................................................................... 3 2.2 Evaluation of the Structural Design ............................................................................... 4 2.3 Evaluation Findings ......................................................................................................... 5 3.0 THERMAL EVALUATION ................................................................................................. 5 3.1 Description of the Thermal Design ................................................................................. 6 3.2 Evaluation of the Thermal Design .................................................................................. 6 3.3 Evaluation Findings ......................................................................................................... 6 4.0 CONTAINMENT EVALUATION ........................................................................................ 6 5.0 SHIELDING REVIEW ........................................................................................................ 6 5.1 Description of Shielding Design ..................................................................................... 7 5.2 Radiation Source Specification ...................................................................................... 7 5.3 Shielding Model ................................................................................................................ 9 5.4 Shielding Evaluation ...................................................................................................... 10 5.5 Evaluation findings ........................................................................................................ 11 6.0 CRITICALITY SAFETY EVALUATION ........................................................................... 12 REFERENCES ........................................................................................................................... 12 CONDITIONS ............................................................................................................................. 13 CONCLUSION ............................................................................................................................ 14
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-9341 Model No. BRR Certificate of Compliance No. 9341
SUMMARY
By application dated April 19, 2018 (Orano, 2018a), as supplemented on September 26, 2018 (Orano, 2018b) and April 4, 2019 (Orano, 2019), Orano Federal Services LLC (Orano or the applicant) requested a one-time authorization for a shipment using the Model No. BEA Research Reactor (BRR) package, Certificate of Compliance (CoC) No. 9341, to transport irradiated (spent) fuel segments in closed-end tubed loaded into up to four payload containers.
The contents will be transported from Oak Ridge National Laboratory (ORNL) to Idaho National Laboratory (INL) under Title 10 of the Code of Federal Regulations (10 CFR) CFR Part 71 in an exclusive use transport.
The applicant requested to use Revision 7 of the CoC as the basis of this request (Orano, 2019). The irradiated fuel segments and the payload containers are not currently authorized contents in CoC No. 9341, Revision 7 (NRC, 2019). Each payload container will be loaded inside the currently approved square fuel basket (AREVA, 2016 and NRC, 2016). The NRC staff (the staff) used guidance provided in NUREG-1617, "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel," for this review.
A one-time letter authorization has been granted to authorize this shipment based on the statements and representations in the application (Orano, 2018a and Orano, 2018b). The staff agrees that the change does not affect the ability of the package to meet the requirements of 10 CFR Part 71.
EVALUATION The applicant requested a one-time authorization for a single shipment of the Model No. BRR package containing irradiated (spent) fuel segments originating from commercial power reactors in the U.S. The staff reviewed and evaluated the fuel parameters provided in the application (Orano, 2018a and Orano, 2018b). The staff determined that the request does not affect the ability of the package to meet the requirements of 10 CFR Part 71. The following sections summarize the staffs evaluation.
1.0 GENERAL INFORMATION 1.1 Packaging The BRR package body is a right circular cylinder 77.1 in. long and 38 in. in diameter. It comprises inner and outer shells connected by a thick lower end casting. The shells and lower end casting are made of American Society for Testing and Materials (ASTM) Type 304 stainless
2 steel with an encased lead shield. The cast-in-place lead shielding fills the annulus between the shells. Together, with the removable 11.2-in. thick shield plug under the closure lid, the package body assembly constitutes the payload cavity, which has a 16-in.
diameter and a 54-in. length.
The principal components of the BRR are:
- a. a lead-shielded package body,
- b. a separate, removable upper shield plug,
- c. a bolted closure lid,
- d. upper and lower impact limiters containing polyurethane foam, and
- e. various payload baskets specifically designed for each type of fuel being transported.
The package is primarily a welded structure using Type 304 austenitic stainless steel. There are no changes in the design of the package body or the payload container. Drawing No.
1910-01-01, Package Assembly SAR Drawing, Revision 4, referenced in CoC, Revision 5, (NRC, 2016 and AREVA 2016) provides the details of the structural design of the package body assembly. A revision to Drawing No. 1910-01-01 was approved in Revision 7 of the CoC (NRC, 2019).
The applicant will transport the fuel segments in closed-end tubes loaded into up to four payload containers. Each payload container is loaded inside the currently licensed square fuel basket
[Assembly A5 on Drawing No. 1910-01-03-SAR, BRR Package Fuel Baskets SAR Drawing, Revision 6, referenced in CoC, Revision 5 (NRC, 2016 and AREVA, 2016)] for transport. The following sections include the evaluation of the changes to the packaging to accommodate the proposed contents requested in this letter authorization.
1.2 Contents The Model No. BRR will be used to ship fuel segments contained in closed-end tubes in up to four containers. The fuel segments have the following characteristics:
- a. contain commercial uranium dioxide (UO2) only,
- b. meet Type B material content,
- c. include fuel discharged from 1995 to 2004, and
- d. fuel rods diameter (including cladding) of approximately 0.4 in.
The applicant provided the bounding parameters of the irradiated fuel segments in its application. Initial enrichments of the fuel segments vary between 2.9 to 4.2 weight percent (wt.%) of Uranium -235 (235U), and burnups varying from 48 to 70 megawatts per day per kilogram of uranium (MWd/kgU), respectively.
3 The rods are approximately 0.4 in. in diameter and vary in length from 1 in. to approximately 13 in. placed in closed-end tubed loaded into up to four payload containers. The total length of the segments will be less than 144 in. (Orano, 2018b). The decay heat generation of the payloads shipment is less than 3 watts (W). These spent fuel rod segments or payload containers are not currently authorized for shipment in the BRR package as approved in CoC No. 9341, Revision 7 (NRC, 2019). As such, the analyses performed for this letter authorization derive their bases from the analyses performed as part of the CoC, Revision 5, of the Model No.
BRR approval (AREVA, 2016), as well as the information provided in the application requesting a letter authorization to ship irradiated fuel (Orano, 2018a and Orano, 2018b). Table 1.2 of this SER summarizes the parameters related to the contents requested in the letter authorization.
Table 1.2. Content parameter limits of the Model No. BRR 2018 letter authorization for transporting fuel segments. (Orano, 2018a and Orano, 2018b)
Parameter Minimum Maximum Units Total fissile mass --- 10.41 grams (g)
(g non-fissile mass)/
Total non-fissile mass 200 ---
(g fissile mass)
Initial enrichment of uranium-235 (235U) 2.9 4.2 wt.%
Mass of 235U --- 3.14 g Mass of plutonium-239 (239Pu) --- 6.30 g Mass of plutonium-241 (241Pu) --- 0.97 g Payload --- 20 pound (lbs.)
Burnup 48 70 MWd/kgU Fuel rod length 1 13 in.
Total length of segments 1 144 in.
Maximum diameter for fuel rod
--- 0.4 in.
(including cladding)
Fuel mass per rod 1.78 2.23 kg Decay heat generation --- 3 W 2.0 STRUCTURAL EVALUATION The purpose of this evaluation is to verify that the BRR transportation package provides adequate protection against loss or dispersal of the proposed radioactive contents as required in 10 CFR Part 71 under normal conditions of transport (NCT) and hypothetical accident conditions (HAC). The following sections document the staffs evaluation of the proposed design changes to the BRR transportation package.
2.1 Description of Structural Design The BRR package is a shipping container to transport irradiated research reactor fuels. The major structural components of the BRR package are the following:
- a. package body
- b. impact limiters
- c. baskets for fuel
4 The applicant proposed transporting in the BRR package the contents described in Section 1.2 of this SER. The applicant did not propose changes to the design of the major structural components of the BRR package.
2.2 Evaluation of the Structural Design The applicant referenced Revision 5 of the CoC for the Model No. BRR in its application as the basis of its initial request. Revision 5 of the CoC authorized the square basket as part of the design for the Model No. BRR (NRC 2016). The applicant would use the square basket for transporting the irradiated fuel segments described in the application for the letter authorization (Orano, 2018a and Orano, 2018b). The NRC staff issued a SER for the BRR package in 2016, corresponding to Revision 5 of the CoC for the Model No. BRR, in which the NRC staff concluded that the BRR package provides adequate structural, thermal, containment, shielding, and criticality protection under NCT and HAC and that the package met the requirements of 10 CFR Part 71 (NRC, 2016).
2.2.1 Materials of Construction The staff performed a review to determine the adequacy of the materials of construction of the payload container, which will house the irradiated pressurized water reactor fuel rod segments.
(Attachment A, Safety Analysis of the Fuel Rod Segment Payload, of the application includes a description of the payload container.) The payload container is loaded inside the currently licensed square fuel basket [Assembly A5 on drawing 1910-01-03-SAR (AREVA, 2016)] for transport. The rod segment/payload container is an American Society for Testing and Materials (ASTM) austenitic stainless-steel weldment not designed as a leak tight vessel. However, the applicant states that the welded construction and dust seal of the container will prevent loss of any material into the package interior during transport.
The staff notes that the applicant does not identify the payload container as important to safety.
Therefore, the applicant did not evaluate the structural criteria or material properties associated with the payload container.
2.2.2 Heavy Loads The applicant stated that the material of the irradiated nuclear fuel rod segment (payload) for a shipment will be secured in up to two identical containers. These containers consist of a section of tubing (ASTM Type 304 stainless steel) with a plate at the closed end, a bolted lid, and a dust seal. The lid is attached with four captured screws and features a pintle for handling in a hot cell. The lower closure plate and the top lid have a nominally 3.25-in. square cross section, and the total length of the container assembly is 39 in. The square fuel basket accommodates the payload containers for shipment in the BRR package.
2.2.3 Conclusion The applicant stated that there are no structural concerns. The staff concludes that no additional structural analysis is required because of the following reasons:
(1) the applicant does not take any credit of the payload container for the shielding analysis, (2) the package has a leak tight design,
5 (3) the welded construction of the payload container will prevent loss of any material into the BRR package cavity during transport, (4) the fuel rod segments will be placed in closed-end tubes prior to placement in the payload container, (5) each payload container has a weight of approximately 20 lb., which is less than the bounding weight of a fuel element in the square fuel basket of 48 lb. in the currently approved BRR design, (6) the payload container can be fit well into the square basket, without risk of interference under NCT and HAC, and (7) the payload container and the package will be loaded dry, which should decrease the possibility of radiolysis or other significant gas generation within the container.
The staff reviewed the application, and its supplements, and finds that currently approved evaluations of the BRR package bound the proposed contents of this request because the following reasons:
(1) the proposed payload has a lighter weight than the bounding payloads, and (2) the payload occupies only up to two cavities in the currently approved square fuel basket.
Also, based on items (1) to (7) of this section, contamination of the interior of the BRR package is not of concern. Therefore, the applicants evaluations are acceptable.
2.3 Evaluation Findings The staff reviewed documentation and statements provided by the applicant to verify that these were within acceptable engineering practices. Based on the review of the statements representations and supplemental information in the application, the staff has reasonable assurance that the applicants request for using the BRR package for a one-time authorization of a shipment of irradiated nuclear fuel rod segments, as described in Section 1.2 of this SER, is adequately described and evaluated to demonstrate that its structural capabilities meet the regulatory requirements of 10 CFR Part 71.
3.0 THERMAL EVALUATION The purpose of this evaluation is to verify that the BRR transport package provides adequate protection for the proposed content against the thermal tests specified in 10 CFR Part 71 and that the package design meets the thermal performance requirements of 10 CFR Part 71 under NCT and HAC.
6 3.1 Description of the Thermal Design The letter authorization proposes a change in the approved contents only, the packaging design features and codes and standards have not changed. The applicants licensing strategy for the heat load specification for the new approved contents was to demonstrate that the new contents were bounded by the heat loads of previously approved contents.
3.2 Evaluation of the Thermal Design The staff reviewed the adequacy of the package materials of the Model No. BRR. The heat load of the proposed content is 3 W. The maximum heat load of the square fuel basket per compartment is 30 W (NRC, 2019). Given the extremely low heat load (3 W) of the proposed content, the thermal analyses corresponding to Revision 7 of the CoC for the Model No. BRR bound the proposed content for the letter authorization. The staff does not have a concern about the temperatures of the contents or the package approaching safety limits for the temperature-sensitive package components (e.g., seals). Therefore, this shipment is authorized with respect to the thermal performance of the package.
The staff reviewed the package description and evaluation and concludes that these satisfy the thermal requirements of 10 CFR Part 71.
3.3 Evaluation Findings Based on its review of the statements and representations in the application, the staff concludes that the applicant adequately described and evaluated the thermal analysis of the proposed content and that the shipment of the proposed content will meet the requirements of 10 CFR Part 71.
4.0 CONTAINMENT EVALUATION The applicant did not propose any changes to the containment evaluation of the BRR package.
Per chapter 4, Containment, of the application, the package is loaded and sealed in accordance with the applicable operating procedures. The package meets the leak tight definition in ANSI N14.5, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, (AREVA, 2018). Therefore, the staff is not concerned about a release of radioactive contents from the package and this shipment would be authorized with respect to the containment performance of the package.
5.0 SHIELDING REVIEW The objective of this review is to verify that the shielding design of the BRR package provides adequate protection against direct radiation from its contents and that the package design meets the external radiation requirements of 10 CFR Part 71 under NCT and HAC.
7 5.1 Description of Shielding Design 5.1.1 Features The BRR package design features consist of a lead-filled shield plug (i.e., top plug), lead-filled package (i.e., side wall), and lead filled bottom (i.e., bottom). These features are briefly described as follows:
(1) Top Plug. The top plug consists of approximately 9.5 in. of lead with a 1-in.
stainless steel bottom cover plate and 0.5-in. stainless steel top plate. The stainless steel lid is 2 in. thick.
(2) Side Wall. The lead in the side wall of the package is 8 in. thick. Also, in the side wall, the inner steel shell is 1-in. thick and the outer stainless steel shell is 2 in. thick.
(3) Bottom. The package bottom consists of a 7.7-in. lead centerline, with a 1-in.
stainless steel bottom cover plate, and approximately 1.2-in. stainless steel inner forging.
Figures 4-1 and 4-2 of the application show an external view and a cross-section view, respectively, of the BRR package. Licensing Drawing No. 1910-01-01-SAR, sheets 2 and 3, show the general structural layout and dimensions of the package body.
5.1.2 Summary of Maximum Radiation Levels Table 6-6 and Table 6-7 of the application includes the summary of the maximum dose rates for NCT and HAC, respectively. Table 5.1 of this SER includes the dose rates calculated by the applicant for NCT.
Table 5.1. Dose Rates of the Model No. BRR with the Proposed Content Under NCT Measurement Location Dose Rate (mrem/hr)
At the Package Surface 49.4 At the Vehicle Surface 23.2 At 2 meters (m.) from the Vehicle Surface 0.3 At the Occupied Location 4.95 x 10-2 For HAC, the applicant calculated a maximum dose rate of 8.6 mrem/hr at one meter from the package surface.
5.2 Radiation Source Specification The applicant evaluated the shielding performance of the BRR package during transport of the spent fuel payload using the following softwares:
- a. SCALE, version 6.2.1. SCALE, version 6.2.1, (SCALE 6.2.1) is used to model the decay of the payload contents and generate gamma and neutron source terms. SCALE 6.2.1 is also used to determine the fissile mass content of the payload.
8
The applicant generated the spent fuel payload source term by following these two steps:
- a. Performed decay modeling to determine the current radioactive isotope inventory of the spent fuel payload (as of January 1, 2018).
The spent fuel payload is composed of three radioactive isotope sets corresponding to payload fuel segments, labeled "605", "616", and "649/650."
Tables 4-1, 4-2, and 4-3 of the application show radioactive isotope sets "605",
"616", and "649/650," respectively. Set "605" is decayed 17.67 years, set "616" is decayed 9.42 years, and set "649/650" is decayed 10.67 years.
- b. Combined all isotope sets to generate gamma and neutron source spectrums.
The applicant multiplied sets "616" and "649/650" by factors of 2.67 and 1.33, respectively, to account for possible additional lengths of fuel beyond that described in the previous Section.
The staff finds the calculations provided by the applicant acceptable mainly because ORIGEN is a widely-used depletion code that has been previously found acceptable by the staff and includes source contribution from radioactive daughter products. The staff performed confirmatory analyses based on the data provided by the applicant and found that the staffs results were in agreement with the applicant results. Sections 5.2.1 and 5.2.2 of this SER includes the staffs evaluation related to the radiation sources considered in the applicants analysis.
5.2.1 Gamma Source For gamma source terms, the applicant performed calculations using the ORIGEN module in SCALE 6.2.1, code package along with an ENDF/B-Vll.1 19-group energy distribution. The gamma source term includes modeling of Bremsstrahlung and daughter products. The applicant properly accounted for secondary gamma radiation in the MCNP shielding calculation.
The staff found that the applicant appropriately calculated the maximum gamma source strength and spectra based on the data provided in the application. Table 4-4 of the BRR fuel rod segment shielding analysis (Orano, 2018a) includes the gamma source spectrum.
The staff performed a confirmatory analysis using ORIGEN-ARP within the SCALE, version 6.1, (SCALE 6.1) depletion code based on the data provided by the applicant and found that the staffs results were in agreement with the applicants results.
5.2.3 Neutron Source For neutron source terms, the applicant performed neutron source calculations using the ORIGEN module in SCALE 6.2.1, code package with an ENDF/B-Vll.1 28-group energy distribution. The applicant did not account for neutrons from subcritical multiplication due to the low mass of fissile material. The staff found that the applicant appropriately calculated the maximum neutron source strength and spectra based on the data provided in Table 4-5 of the BRR fuel rod segment shielding analysis (Orano, 2018a), which includes the neutron source spectrum.
9 5.3 Shielding Model The applicant states that for the shielding model, the assumptions are as follow:
- a. The vehicle width is assumed to be 8 feet (ft.).
- b. The package lead radial shrinkage gap is 0.0625 inches and the package lead axial gap is 1.18 inches.
- c. The occupied location is assumed to be 25 ft. from the centerline of the package.
- d. All spent fuel and fuel basket geometry is conservatively ignored.
(1) The spent fuel source is modeled as a point source within a 1-centimeter (cm) diameter sphere.
(2) The source is able to move to any location within the package inner cavity, bounding any possible reconfiguration of the spent fuel during NCT or HAC.
(3) The spent fuel will be contained within a robust, sealed container.
(4) All segments will be contained in closed-end tubes and up to four containers.
(5) The container will be constrained within one of the openings of the square fuel basket.
- e. Subcritical multiplication is not accounted for. (The package payload is exempt from criticality analysis.)
The staff finds these assumptions acceptable because of the following reasons:
- a. The width is a standard width of a trailer.
- b. The packages lead radial shrinkage gap and the lead axial gap increases the streaming path for gamma rays resulting in conservative dose rates calculations.
- c. A reasonable assumption of the occupied location based on the transportation configuration.
- d. A point source geometry is conservative because it does not account for any self-shielding from the source.
- e. The subcritical multiplication does not have a significant effect on dose rates.
The applicant states that the spent fuel payload consists of fuel segments between 1.0 and 13 in. in length. The spent fuel segments have a 0.4-in outer diameter. The spent fuel payload will be transported using the square fuel basket shown in Figure 4-3 of the BRR fuel rod segment shielding analysis (Orano, 2018a), which includes the gamma source spectrum. An additional
10 20 in. of set "616" and 9 inches of set "649/650" may be included in the payload (Tables 4-1 and 4-3 respectively show the customer-provided radioisotope sets corresponding to the total aggregate fuel lengths of 12 in. for "616" and 27 in. for "649/650"). The applicant provided the parameter used to model the fuel segments radioactive isotope set in document its response to the staffs request for additional information (Orano, 2018a).
The applicant performed the shielding analysis of the package using MCNP, version 6.1, with continuous energy ENDF/B-VII neutron and photon cross section libraries. The applicant explicitly used the energy distribution of the source term calculated in the MCNP model and performed separate calculations for each of the three source terms (i.e., primary gamma dose rates, neutron dose rates, and secondary gamma dose rates). The applicant modeled the source for the spent fuel as a point source to determine the five worst-case locations where streaming path can occur with three runs per location. The staff found this approach acceptable for bounding any possible reconfiguration of the spent fuel during NCT or HAC. The applicant multiplied sets "616" and "649/650" are multiplied by factors of 2.67 and 1.33, respectively, to account for possible additional lengths of fuel. The five source locations within the package cavity are the following:
- a. source shifted to the lower wall and radial wall ("Bottom" case);
- b. source shifted to the lower wall and aligned with the drain port ("Drain" case);
- c. source shifted to the radial wall in the axial center of the cavity ("Middle" case);
- d. source shifted to the upper wall and radial wall ("Top" case); and
- e. source shifted to the upper wall and aligned with the vent port ("Vent" case).
These runs included primary gamma dose rates, neutron dose rates, and secondary gamma dose rates. The modeled package is presented in Figure 5-1 of the BRR fuel rod segment shielding analysis (Orano, 2018a). The applicant modeled all relevant package features in three-dimensions and the impact limiters as void. All space outside of the package is modeled as void space. The staff finds these modeling assumptions acceptable because modeling in three-dimension can provide accurate results based on previous analysis and obtain uniformly small relative uncertainties. For NCT, the applicant included both the neutron shield and impact limiters in the package model. For HAC, the applicant replaced the neutron shield with void and completely removed the impact limiters. Using void in the model is conservative because the components modeled as void space provide shielding to the system. The staff determined that all reconfiguration of the package internal cavity during an accident condition was bounded by the evaluated worst-case locations due to the non-physical concentration of the spent fuel source terms.
5.4 Shielding Evaluation The applicant utilizes several assumptions throughout the shielding calculations to provide assurance that the actual dose rates remain below the regulatory limits.
For NCT, the applicant used five mesh tallies at different locations of the package to measure the surface dose rates. Those locations are as follows:
- a. the package side surface,
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- b. the top surface of the top impact limiter,
- c. the bottom surface of the bottom impact limiter,
- d. the impact limiter side surfaces, and
- e. the impact limiter ' underside' surfaces.
The applicant considered the impact limiter underside surfaces as part of the package side surface. Also, the applicant considered two segmented surface tallies on the conical surfaces of the top and bottom impact limiters in the package modeling. These conical surfaces are part of the top and bottom surfaces of the package. Vehicle surface dose rates are measured using one mesh tally at the projected transport vehicle side surface (4 ft. from package centerline).
The vehicle top surface is the same as the package top surface, while the vehicle bottom surface is conservatively measured at the package bottom surface. Two-meter dose rates are measured using one mesh tally at 2 m. from the transport vehicle projected side surface. The staff finds the inclusion of the impact limiters in the shielding analysis for NCT acceptable.
For HAC, dose rates at 1 meter are measured using three mesh tallies at the following locations:
- a. 1 meter from the package side,
- b. 1 meter from the package top surface, and
- c. 1 meter from the package bottom surface.
For HAC dose rates, the applicant takes no credit for the geometry of the impact limiters, bounding the crush of the impact limiters during an accident. The staff finds this approach acceptable because their omission should, in all cases, be conservative.
MCNP uncertainties, as a statistical code, are expressed as the standard deviation of the mean.
The applicant chose parameters, such as variance reduction and the number of starting particles for each run, to obtain a relative error of less than 3.5% for the dose rates associated with HAC. The staff finds this method acceptable since the main reason to use variance reduction in the shielding calculations is to calculate fluxes and dose rates with low uncertainties.
Tables 6-6 and 6-7 of the application include the external dose rates on the surface of the package are presented for NCT and HAC, respectively.
5.5 Evaluation findings The staff reviewed and evaluated the description of the package design features related to shielding and the source terms for the design basis fuel. The methods used are consistent with accepted industry practices and standards. The staff reviewed the maximum dose rates for NCT and HAC and determined that the reported values were below the regulatory limit in 10 CFR 71.47 and 71.51.
12 Based on its review of the statements and representations provided in the application, the staff has reasonable assurance that the shielding evaluation is consistent with the codes, standards, and NRC guidance applicable to the package shielding analyses, and that the package design and contents satisfy the shielding and dose limits required in 10 CFR Part 71.
6.0 CRITICALITY SAFETY EVALUATION The following table includes the maximum mass of fissile material in the Model No. BRR:
Isotope Mass (g) 235U 3.14 239Pu 6.30 241Pu 0.97 The total mass of fissile material associated with this shipment is 10.41 g. Therefore, as it pertains to this letter authorization, the package is fissile exempt per 10 CFR 71.15, since the total mass of fissile material per shipment is less than 15 g.
REFERENCES (ANSI, 1997) American National Standards Institute ANSI N 14.5-1997, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, ANS, LaGrange Park, IL.
(AREVA, 2016) Noss, Phil, AREVA Federal Services LLC (AREVA), letter to U. S.
Nuclear Regulatory Commission (NRC) (Attn: Document Control Desk), May 26, 2016, ADAMS Accession No. ML16168A266).
(AREVA, 2018) Noss, Phil, AREVA Federal Services LLC (AREVA), letter to U. S.
Nuclear Regulatory Commission (NRC) (Attn: Document Control Desk), January 30, 2018, ADAMS Package Accession No.
(NRC, 2016) McKirgan, John, U. S. Nuclear Regulatory Commission (NRC),
letter to Philip W. Noss, AREVA Federal Services LLC (AREVA),
July 21, 2016, ADAMS Package Accession No. ML16204A306.
(NRC, 2019) McKirgan, John, U. S. Nuclear Regulatory Commission (NRC),
letter to Philip W. Noss, Orano Federal Services LLC (Orano),
February 5, 2019, ADAMS Package Accession No.
(NUREG-1617) U.S. Nuclear Regulatory Commission, Standard Review plan for Transportation Packages for Spent Nuclear Fuel, NUREG-1617, Initial report, March 2000, ADAMS Accession No. ML003696262.
13 (Orano, 2018a) Noss, Phil, Orano Federal Services LLC (Orano), letter to U. S.
Nuclear Regulatory Commission (NRC) (Attn: Bernie White), April 19, 2018, ADAMS Accession No. ML18122A065.
(Orano, 2018b) Noss, Phil, Orano Federal Services LLC (Orano), letter to U. S.
Nuclear Regulatory Commission (NRC) (Attn: Norma Garcia-Santos), September 26, 2018, ADAMS Accession No.
(Orano, 2019) Noss, Phil, Orano Federal Services LLC (Orano), letter to U. S.
Nuclear Regulatory Commission (NRC) (Attn: Norma Garcia-Santos), April 4, 2019, ADAMS Accession No. ML19100A111.
(SCALE 6.1) Scale 6.1, A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, (includes ORIGEN).
(Released on January 31, 2014).
CONDITIONS CoC No. 9341 has been amended by letter to authorize one shipment of the Model No. BRR package containing irradiated fuel segments. The following conditions apply:
- 1. The fuel assemblies will be shipped by truck.
- 2. There will be a maximum of one BRR package per shipment on a single truck.
- 3. The fuel assemblies will only contain commercial UO2.
- 4. The fuel assemblies will meet Type B material contents.
- 5. The shipment will only include irradiated (spent) fuel discharged from 1995 to 2004.
- 6. The fuel rods diameter (including cladding) is approximately 0.4 in.
- 7. The following table includes additional specific limits about the contents:
Parameter Minimum Maximum Units Total fissile mass --- 10.41 g (g non-fissile mass)/
Total non-fissile mass 200 ---
(g fissile mass)
Initial enrichment of235U 2.9 4.2 wt.%
Mass of 235U --- 3.14 g Mass of 239Pu --- 6.30 g Mass of 241Pu --- 0.97 g Payload --- 20 lbs.
Burnup 48 70 MWd/kgU Fuel rod length 1 13 in.
Total length of segments 1 144 in.
Maximum diameter for fuel rod
--- 0.4 in.
(including cladding)
Fuel mass per rod 1.78 2.23 kg Decay heat generation --- 3 W All other conditions of CoC No. 9341, Revision 7, shall remain the same (NRC, 2019). This authorization expires on September 1, 2020.
14 CONCLUSION CoC No. 9341 has been amended by letter to authorize one shipment of irradiated fuel segments in the Model No. BRR package. This authorization expires on September 1, 2020.
Based on the statements and representations in the application, and with the conditions listed above, the staff agrees that this change does not affect the ability of the package to meet the requirements of 10 CFR Part 71.
Issued on 6/14/19.