ML21140A092
| ML21140A092 | |
| Person / Time | |
|---|---|
| Site: | 07109792 |
| Issue date: | 05/26/2021 |
| From: | Plate N US Dept of Energy, National Nuclear Security Admin |
| To: | John Lubinski Office of Nuclear Material Safety and Safeguards |
| References | |
| NR:RR:NSPlate G#21-01990 | |
| Download: ML21140A092 (15) | |
Text
DEPARTMENT OF ENERGY NATIONAL NUCLEAR SECURITY ADMINISTRATION 1000 INDEPENDENCE AVENUE SW WASHINGTON DC 20585*1000 NR:RR:NSPlate G#21-01990 April 26, 2021 John Lubinski Director, Office of Nuclear Material Safety and Safeguards Nuclear Regulatory Commission Washington, DC 20555 MODEL 1 D1 G CORE BASKET THERMAL SHIELD SHIPPING AND STORAGE CONTAINER - SAFETY ANALYSIS REPORT FOR PACKAGING TO SUPPORT DISPOSAL SHIPMENT OF D2W CORE BASKET; RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION
Reference:
(a) NR letter G#C20-04104 dated October 13, 2020 (b) NNL meeting summary RSS-SC-SCA-00390 dated March 17, 2021
Background:
The Model 1 D1 G Core Basket - Thermal Shield Shipping and Storage Container (herein referred to as the Model 1 container) is a single-use container originally designed to transport irradiated D1 G core basket thermal shield assemblies to a low-level radioactive waste facility for disposal. The Model 1 container is certified as a Type B package for shipment of highly radioactive material. Title 10, Code of Federal Regulations, Part 71 (1 0CFR71) delineates federal requirements for Type B packages.
In the reference (a) letter, Naval Reactors forwarded to the NRC for review and concurrence an addendum to the Model 1 container Safety Analysis Report for Packaging (SARP) to support shipment of irradiated D2W core baskets. Naval Reactors, NRC, and Naval Nuclear Laboratory conducted several productive meetings to discuss early NRC questions on the SARP, as documented in the reference (b) meeting summary. At the conclusion of these meetings, all parties agreed that several NRC questions required formal documentation to disposition. NRC provided a request for additional information to Naval Reactors in a letter dated March 12, 2021.
Naval Reactors Action: Enclosure (1) to this letter provides responses to the NRC request for additional information. Naval Reactors appreciates the NRC's thorough and timely review efforts of the application and if you have any additional questions, please do not hesitate to call me at (202) 781-6034.
- 1/j~
N. S. Plate Naval Reactors Enclosure and Copy to: see next page.
2 NR:RR:NSPlate G#21-01990
Enclosure:
(1) NAVAL REACTORS RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION (DOCKET NO. 71-9792)
Copy to:
KAPLADSARS NRLFO NRLFO-Transportation Division [M. Salamon]
Director, Division of Fuel Management, NMSS, NRC [A. Kock]
Chief, Storage and Transportation Licensing Branch , DFM, NMSS, NRC [J . McKirgan]
Project Manager, STLB, DFM, NMSS, NRC [C. Allen]
General Manager, NNL [G. Lubinsky]
Manager, Reactor Servicing, NNL [V. Pantloni]
Manager, Reactor Servicing Systems, RS, NNL [M. Drewen]
Manager, Shipping Containers, RSS, RS, NNL [S. Fiscus]
Manager, Shipping Container Analysis, SC, RSS, RS, NNL [C. Haslett]
ENCLOSURE (1)
NAVAL REACTORS RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION (DOCKET NO. 71-9792)
Enclosure (1) to s*er 08G#21-01990
Request for Additional Information Docket No. 71-9792 Model No. D1 G Core Basket Thermal Shield Shipping and Storage Container Package This enclosure restates the NRC Request for Additional Information provided via letter dated March 12, 2021 and provides Naval Reactors responses.
Chapter 3 - Thermal Evaluation 3.1 Justify the maximum outer surface temperature reported in the safety analysis report (SAR) for the hypothetical thermal accident condition.
The SAR Thermal chapter reports an outer package surface temperature that is much less than the 1475°F flame temperature imposed for the hypothetical accident condition (HAC) fire even though, according to the simple pictorial representation, the outer surface appears to be exposed to the flame. In addition, previous applications for this package calculated an external package temperature much closer to 1475°F. Staff noted that sensitivity calculations by the applicant have analyzed the effect of puncture-damaged containers, whereby the external temperature is close to the 1475°F, and the exposed damaged sections that are indented relative to the undamaged outer surface have temperatures much less than the flame temperature. Staff further noted that the sensitivity calculation neither explicitly described the boundary condition at the rounded puncture-damaged surface (compared to the boundary condition of the undamaged outer package surface) nor discussed if the reduced temperature at the rounded puncture-damaged surface was due to the radiation heat transfer view factors or other effects. If the current SAR thermal analysis remains unchanged as a result of this RAI, provide additional supporting information as to why package temperatures are representative of the applied hypothetical accident fire conditions in accordance with 10 CFR 71.73, and if the SAR thermal analysis is updated as a result of this RAI, discuss potential effects to the packaging if external temperatures approach 1475°F. The response also should provide aspects of previously approved Docket Number 71-9792 package applications (e.g., May 1986 application) that demonstrate the current package is bounded by those results.
This information is needed to verify compliance with Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 71 .51 and 10 CFR 71. 73.
Response: The package detailed in the original Docket Number 71-9792 analysis with the D1 G-2 core barrel, as revised by the May 1994 application, has similar thermal characteristics compared to the package analyzed in the current application with the D2W core barrel. Both packages are of similar weight, are Enclosure (1) 2
constructed of similar materials, and have identical heat-transfer paths to the interior of the package. The only thermal difference between the packages is the higher cargo decay heat in the D1 G-2 package. However, cargo decay heat is not a significant driver of package temperature compared to the HAC fire heat input.
The most-recent D1 G-2 analysis (May 1994 application) reported package surface temperatures approaching 1,475°F during the HAC fire. Given an identical set of assumptions, the D1 G-2 package thermal analysis conclusions would bound that of the D2W package detailed in the current application. However, the D2W thermal analysis uses different model assumptions to reach a more limiting internal temperature and resulting pressure, driven by the fire input energy.
The fire heat input is thermally modeled with radiative and convective heat transfer on the applicable outer surface discussed below. The surface temperature is dependent on the rate of fire heat applied and the rate that the heat energy diffuses into the container. Figure 1 depicts a comparison of the SARP thermal analysis for the D2W package and for the D1 G-2 package. The major difference between the analysis assumptions for the D2W package and the D1 G-2 package is the location where the fire is applied. The D2W analysis applies the fire to the outside of the thick steel wall of the inner vessel that has a high thermal diffusivity ,
while the D1 G-2 analysis applies the fire to the thin steel outer wrapper that is backed by concrete that has a large thermal resistance. This results in heat rapidly migrating into the container for the D2W analysis whereas the heat transfer in the D1 G-2 analysis encounters significantly more resistance after the thin steel outer wrapper. Accordingly, the surface temperature in the D2W analysis is less than the surface temperature in the D1 G-2 analysis during the fire transient.
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Enclosure (1) 3
As shown in Figure A.3.4-2 (page A.3.4-6) of the D2W SARP thermal analysis, the outer surface temperature of the inner vessel rapidly increases, approaching the applied fire temperature, during the transient. However, due to the large thermal mass and high thermal diffusivity of the inner vessel, the surface temperature only reaches a maximum temperature of 759°F before the fire input is removed after 30 minutes. While localized regions of outer package surfaces may approach the fire temperature of 1475°F, exterior package temperatures are not the primary focus of the thermal analysis. The melting point of the outer wrapper and inner vessel steel is roughly 1,000°F greater than the fire temperature, thus peak package temperature is not a structural failure concern.
The assumption change between the analyses is the package surface receiving the fire boundary condition. The change ensures a conservative and limiting internal temperature and resulting pressure used to demonstrate package containment. The red lines in Figure 1 denote the inside surface of the sealed inner vessel. The temperature at this location feeds into calculation of a conservative maximum internal pressure, which is required to demonstrate package containment.
Additional Discussion on Thermal Analysis Assumptions and Results In the D1 G-2 analysis, the fire energy must diffuse through a high thermal resistance layer prior to reaching the inner vessel. Figure 1 shows nine inches of concrete, but the thermal analysis neglects the mass of the concrete due to the possibility of concrete damage during the HAG free drop and puncture conditions and models this layer with only radiative heat transfer (layer effectively modeled as vacuum). This removes any availability for conduction of fire energy through the concrete layer into the container in the D1 G-2 analysis. The concrete layer is not modeled in the D2W analysis; this analysis applies the fire energy directly to the inner vessel to create a bounding heat input condition with respect to the inner surface of the inner vessel, thus maximizing package internal pressure.
As documented in the reference (a) meeting summary, a sensitivity study supporting the D2W SARP thermal analysis was discussed that involved the case in which the package retains the concrete layer and outer steel wrapper with explicit pin puncture damage. Figure 2 depicts the differences between the sensitivity study and the D2W SARP thermal analysis. The sensitivity study explicitly includes the concrete layer with a puncture pin indent in the thermal model. The modeled puncture damage represents a full revolution "slot" due to the nature of the axisymmetric model with the fire applied along all faces of the "slot".
As stated above, the D2W SARP thermal analysis assumes loss of the concrete and wrapper and places the fire against the outside surface of the inner vessel.
Enclosure (1) 4
Concrete, Steel, highly high th rmal conductiv , low Container Interior resistance Contain r Interior thermal resistance Fire Fire Sensitivity Study O2W SARP Analysis Figure 2. D2W Core Barrel Pin Damage Comparison The sensitivity study provides a better comparison to the 01 G-2 SARP analysis than the O2W SARP analysis due to the surface where the fire is applied and the similar thermal resistance path for the fire energy. This path includes high thermal diffusivity through the steel outer wrapper, followed by a large thermal resistance through the concrete layer. The sensitivity study uses concrete conduction, while the 01 G-2 SARP analysis uses radiative heat transfer through the concrete layer (concrete failure assumed). Closed-form calculations show comparable thermal resistance between conduction through the concrete layer and radiative heat transfer over the same layer if the concrete is removed.
Table 1 and Figure 3 show the temperatures in the pin puncture sensitivity analysis and how they compare with 01 G-2 and O2W SARP analyses.
Table 1. Thermal Model Results Maximum Interior Maximum Exterior Model Temperature (°F)* Temperature (°F)
O2W SARP 441 759 Pin Puncture Sensitivity 290 1 437 O1G-2 SARP 276 1,323
- Maximum interior temperature occurs after the end of the 30-minute fire , not pictured below.
Enclosure (1) 5
NT1 1 NT11 14375 15e e 13336 71 4.3 1229.6 669.7 11 25.6 625.1 1021 6 580.6 917.6 536.0 8137 491.4 709.7 446.9 605.7 402.3 50 1 7 '357.7 3977 313.1 2937 268.6 1898 224.0 y
y Pin Damage Model No Concrete Model Figure 3. D2VV Core Barrel Model Results at End of 30-Minute Fire Table 1 and Figure 3 show that the sensitivity model and the D1 G-2 model produce similar results with consistently lower interior temperatures when compared to the D2W SARP model. The higher interior temperature in the D2W SARP analysis resulted in a more limiting pressure in the container, and was therefore used as the D2W SARP analysis of record .
No SARP changes are planned as part of resolution of this question.
Chapter 4 - Containment Evaluation 4.1 Justify the time period to complete shipment that was calculated in the radiolysis analysis.
The radiolysis calculation was based on the number of days to reach a hydrogen flammable limit. Typically, applications for packages have relied on some margin when specifying the time period to complete shipment such that the flammable limit is not reached (e.g. , Chapter 5 in NUREG/CR-6673).
This information is necessary to determine compliance with 10 CFR 71.43(d).
Response: The SARP shipment restriction and associated Certificate of Compliance (CoC) requirement to complete shipment within 1,444 days after the container is sealed ensures the package will not exceed the 5 percent hydrogen Enclosure (1) 6
concentration limit. Conservatisms within the hydrogen generation analysis provide margin to the 5 percent limit, vice applying margin in the shipment duration restriction.
The hydrogen generation rate analysis applies the peak gamma energy deposition rate per unit volume of water at any point on the inner thermal shield to all of the residual water permitted in the container. This is conservative since the volume averaged energy deposition rate in water volume would be lower based on actual water location. In addition, the analysis does not account for reduction in the hydrogen generation rate due to decay of the core basket activation sources after the container is sealed. The most significant longer-lived nuclide contributing to the gamma energy deposition rate is 6°Co, with a half-life of 5.271 years. If the time to reach the 5 percent limit accounted for source decay, the hydrogen concentration limit would be reached at about 2,038 days after sealing the container. Furthermore, the analysis conservatively neglects the presence of catalyst in the container that would reduce hydrogen concentration levels. From an operational perspective, Naval Reactors experience confirms this analysis methodology is conservative as similar methods have been applied to our Program spent fuel containers and post-shipment hydrogen surveys have repeatedly returned hydrogen levels significantly less than calculated in the associated SARPs. Also , given the short travel distance and extensive shipment planning, Naval Reactors expects all shipments would be completed within one year of sealing the container. Based on the conservatisms stated above, Naval Reactors considers that the hydrogen generation rate analysis and associated shipment time restriction provide sufficient confidence that the package will not reach the 5 percent hydrogen concentration limit.
No SARP changes are planned as part of resolution of this question.
4.2 Explain the determination of the hydrogen generation rate used in the radiolysis calculation .
The hydrogen generation rate is a function of the radiolytic G value, the decay heat absorbed by the radiolytic material, etc. There was no explanation provided as to the basis for the specified hydrogen generation rate which is an important parameter when considering impacts within the package.
This information is necessary to determine compliance with 10 CFR 71.43(d).
Response: The hydrogen generation rate is determined by multiplying a radiolytic G-value by the energy deposition rate in residual water within the container. The radiolytic G-value used is 0.45 molecules of H2/100 eV of gamma radiation deposited. This G-value is consistent with the highest G-value for gamma Enclosure (1) 7
radiation deposition in liquid water in Table D.1 of NUREG/CR-6673, and does not account for recombination. The rate of energy deposition in water is determined by multiplying a gamma energy deposition rate per unit water volume by the maximum residual volume of water, and applying a 1.15 method uncertainty factor.
The gamma source distribution in the core barrel assembly is the same as provided in Chapter A5. The gamma energy deposition rate is at the worst-case point location that water could realistically accumulate on the cargo (at the inner radius of the core barrel inner thermal shield). These assumptions yield an energy deposition rate of 2.33 watts deposited in water at 136 days after shutdown. The above justification was included in underlying analysis documents supporting the SARP; however, for completeness, Naval Reactors plans to make the following change to the SARP to explicitly include this technical justification:
Chapter A2 - Section A.2.2.2 (insertions in bold)
"The D2W Package contains a maximum of 3.63 gallons of residual water during shipment. The residual water can be transformed into hydrogen gas by gamma induced radiolytic decomposition. The hydrogen concentration in the package must be limited to less than five percent by volume to avoid a flammable mixture.
The hydrogen generation rate is determined by multiplying a radiolytic G value by the energy deposition rate in the residual water. The radiolytic G value used is 0.45 molecules of H2/100 eV of gamma radiation deposited.
This G value is consistent with the highest G-value for gamma radiation deposition in liquid water in Table D.1 of NUREG/CR 6673 (2000). The rate of energy deposition in water is determined by multiplying a gamma energy deposition rate per unit water volume by the maximum residual volume of water, and applying a 1.15 method uncertainty factor. The gamma energy deposition rate per unit volume is at the worst-case point location that water could realistically accumulate on the surface of the core barrel. These assumptions yield an energy deposition rate of 2.33 watts deposited in residual water at 136 days after shutdown. The following calculation determines the amount of time for the volume of hydrogen in the Model 1 shipping container to reach a five percent concentration (conservatively neglecting gamma source decay after the container is sealed}."
Chapter 5 - Shielding Evaluation 5.1 Provide more details about the program used to determine the activation source.
The applicant states that it determined the source term by calculating a bounding flux and then used a program "similar to" ORIGEN-S to calcu late nuclide concentrations. The applicant provided neither details regarding how the code Enclosure (1) 8
solves for the nuclide vectors, nor similarities and differences of the code to ORIGEN-S.
This information is necessary to determine compliance with 10 CFR 71.33(b)(1) and 71.47(b).
Response: The SPENT3 computer code computes the buildup of radionuclides due to neutron absorption and their decay primarily based on ENDF/B-VII cross section data. SPENT3 was qualified by comparing nuclide activities to results from ORIGEN-S calculations, measured activity in a component irradiated near the fuel in an NNPP reactor plant, and measured after-shutdown radiation levels in an NNPP reactor plant.
MC21 is a Monte Carlo radiation transport code that calculates neutron and photon transport in fixed source problems. MC21 was qualified by comparisons to measurements, Monte Carlo transport calculations performed with the MCNP computer code, and discrete ordinates gamma transport calculations (PARTISN computer code). For completeness, Naval Reactors plans to make the following changes to the SARP to include details related to the computer codes used in the SARP analyses:
Chapter A5 - Section A.5.2.1.1 (insertions in bold, deletions in stril<ethrough)
"Activation gamma source terms are developed using a-the SPENT3 computer program, which is similar to ORIGEN-S (Reference A.5.5-1). ORIGEN Sis a computer program used to perform irradiated material analyses. The main function of the computer programSPENT3 is to solve for the time-dependent concentrations of structural activation nuclides, which are simultaneously generated or depleted through neutron transmutation or radioactive decay primarily based on ENDF/8-VII cross section data. The energy-dependent gamma source levels are computed using first principles by appropriately summing contributions of the individual gamma energies according to the energy group to which they belong.
For gamma source generation, gamma/x-ray energy data is represented in discrete form as pairs of data - the intensity of a gamma and the energy at which the gamma is emitted. The gammas are binned into 27 energy groups ranging from 0.001 to 3.85 MeV. SPENT3 was qualified by comparing nuclide activities to results from ORIGEN-S calculations, measured activity in a component irradiated near the fuel in an NNPP reactor plant, and measured after shutdown radiation levels in an NNPP reactor plant."
Chapter A5 - Section A.5.2.3 (insertions in bold, deletions in stril<ethrough)
"The activity due to neutron activation of the 02W CB/TS assembly is 2716 TBq at Enclosure (1) 9
136 days after shutdown. The radionuclides contributing greater than 3.7E-02 TBq (one curie) are presented in Table A.5.2-4; these radionuclides make up 99.99% of the activity due to neutron activation. The activating neutron fluxes used for the activity calculation are the same used for the gamma source determination, including a 1.5 method uncertainty factor applied to the flux above and below the fuel region. The activation curie content calculations contain no other method uncertainty or design assurance factors. The activity in the Model 1 container is calculated using a-the SPENT3 computer program similar to ORIGE:~J S (Reference 1\.6.6 1). ORIGE:~J Sis a computer program used to perform irradiated material analyses. The main function of the computer programSPENT3 is to solve for the time-dependent concentrations of structural activation nuclides, which are simultaneously generated or depleted through neutron transmutation or radioactive decay primarily based on ENDF/B-VII cross section data. The material compositions used for the activity calculations are provided in Table A.5.3-2."
Chapter A5 - Section A.5.4.1 (insertions in bold, deletions in stril<ethrough)
"The radiation levels from D2W CB/TS assembly activation gamma sources and crud gamma sources deposited on MIL-EN82 cladding surfaces were calculated using the MC21 computer code, a continuous-energy, Monte Carlo transport code (Reference A.5.5-4). Sufficient particle histories are run to achieve peak radiation level results with 95 percent confidence interval relative error (95CI RE) of the mean below 10 percent at all reported locations. MC21 was qualified by comparisons to measurements, Monte Carlo transport calculations performed with MCNP, and discrete ordinates gamma transport calculations (PARTISN). The radiation levels from relocated crud present in the residual water was calculated with the MCNP computer code (Reference A.5.5-5) using a supplemental point-kernel method. The point kernel supplemental method to MCNP is similar in function to the SPAN computer code (Reference A.5.5-6). Iron dose buildup factors are used for all tallies within the MCNP/point kernel hybrid method."
Chapter 9 - Quality Assurance Requirements 9.1 Describe how the 10 CFR 71, Subpart H requirements are addressed as a basis for demonstrating compliance with 10 CFR 71, Subpart H, "Quality Assurance,"
program for use by the Navy Nuclear Propulsion Program (NNPP).
The NRC's regulatory requirements for packaging and transporting radioactive materials are codified in 10 CFR Part 71. Package approval must include a quality assurance (QA) program description or a reference to a previously approved QA program applicable to the package. Further, Subpart H of 10 CFR Part 71 Enclosure (1) 10
establishes QA requirements that apply to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of packaging of components under control of the QA program to ensure the integrity of the packaging and its capability to prevent or mitigate the consequences that could result from release of radioactive material.
The staff reviewed NNPP's submittal of its QA program description for compliance to the related regulatory requirements. The staff noted that NNPP's Safety Analysis Report for Packaging (SARP), Section 9 provides a description of the program covering NNPP's QA program NNPP-SNF-YMSA-19 (Background, Evaluation, and Analysis Report (BEAR) 19, Revision 1), as it pertains to the disposal of naval spent nuclear fuel (SNF) in the geologic repository at Yucca Mountain, Nevada, and that elements of the NNPP QA program comply with the requirements of 10 CFR 63.142, Disposal of High-Level Radioactive Wastes in a Proposed Geologic Repository at Yucca Mountain, Nevada, Subpart G, Quality Assurance Criteria. Additionally, the staff noted that NNPP determined that the NNPP QA program is compatible with the key quality objectives of the Department of Energy Office of Civilian Radioactive Waste Management (DOE-RW) Quality Assurance Requirements and Description (OARD), No. DOE/RW- 0333P.
However, Section 9, did not address what aspects of, or to what extent, the QA program provided will be applicable to the Core Basket-Thermal Shield Shipping and Storage Container package to fulfill the requirements of 10 CFR Part 71 .37 and Part 71 Subpart H. Therefore, the staff needs additional information.
This information is needed to determine compliance with 10 CFR 71.37(a) and 71.101 (b), (c), and (f).
Response: The Naval Nuclear Propulsion Program (NNPP) independently certifies Type B containers to 10CFR71, and subsequently ships them, under authority provided to the Department of Energy in 49CFR173.7.(d). However, over the years, the NNPP has created a standard practice of obtaining NRG technical review and concurrence of NNPP SARPs for the Type B containers the NNPP certifies to 10CFR71. The primary purpose of this process is to ensure the NNPP is appropriately interpreting and applying the NRG regulations in 10CFR71.
The NNPP maintains an independent Quality Assurance (QA) Program via Executive Order 12344, codified in 50 USC sections 2406 and 2511, that is applied to NNPP regulated activities, including Type B container design, analysis, certification, and operation. The NRG does not provide oversight of the NNPP QA Program, with exception to Yucca Mountain activities (discussed below).
Therefore, the NNPP neither requests nor receives NRG approval of its QA program during NRG review of Type B packages. However, the NNPP has historically provided a detailed explanation of how its QA Program satisfies 10CFR71 Subpart H requirements in Chapter 9 (Quality Assurance Requirements)
Enclosure (1) 11
of NNPP SARPs. In light of the process discussed above, the NRC has historically not documented reviews of Chapter 9 in Safety Evaluation Reports for NNPP SARPs. For example, none of the NRC Safety Evaluation Reports for the 9792 package acknowledge Chapter 9.
The original SARP for the 9792 package includes a Chapter 9 that describes the NNPP QA program and how it satisfies requirements in 10CFR71 Subpart H. Note that Chapter 9 in all current NNPP SARPs are largely identical. However, in the two applications currently under NRC review, the NNPP provided reference to Background, Evaluation, and Analysis Report 19, "NNPP Quality Assurance Program" (BEAR 19) vice a detailed description of the NNPP QA Program in Chapter A9.
BEAR 19 describes the NNPP QA Program for the Yucca Mountain Licensing activities, which required NRC oversight and approval. For Yucca Mountain Licensing activities, the NNPP QA Program fell within the overall QA Program of The Office of Civilian Radioactive Waste Management (RW). The NRC reviewed and accepted the RW QA program, Reference (2) (NNPP interface discussed in Section A.2.1 ), as documented in Section 2.5.1 of the Yucca Mountain Safety Evaluation Report Volume 4, Reference (3). The NNPP previously provided BEAR 19 to the NRC as documented in Reference (4). The NNPP intended to use the QA Program outlined as part of these Yucca Mountain Licensing activities to apply to NNPP transportation casks. Transportation casks used for spent fuel shipments to the repository require NRC certification in accordance with the Nuclear Waste Policy Act, including certification of the NNPP QA Program to 10CFR71 subpart H.
However, these Licensing Activities were paused before any NNPP transportation casks went through the certification process. Naval Reactors plans to obtain explicit NRC approval of the NNPP QA Program to 10CFR71 subpart Has part the application for naval spent fuel canisters in the M-290 shipping container, currently expected to be provided to NRC for review in 2029.
The NNPP recognizes that BEAR 19 focuses on satisfying requirements in 10CFR63 and does not provide an explicit connection to the 10CFR71 subpart H requirements. However, BEAR 19 provides the same fundamental information about the structure and requirements of the NNPP QA Program as previously provided in SARP Chapter 9s. NNPP SARPs provide this information for perspective on the NNPP QA Program, although NRC approval of the NNPP QA Program to subpart H is not required to support current NNPP transportation activities. The NNPP will include the reference to BEAR 19 in future applications.
No SARP changes are planned as part of resolution of this question.
Enclosure (1) 12
References:
Reference (1): NNL Meeting Summary RSS-SC-SCA-00390 dated March 17, 2021 Reference (2): DOE/RW-00333P, Quality Assurance Requirements and Description, Revision 20, October 1, 2008 Reference (3): NUREG-1949, Safety Evaluation Report Related to Disposal of High-level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada, Volume 4, Administrative and Programmatic Requirements, December 2014 Reference (4): NR letter U#08-03589 dated September 16, 2008 Enclosure (1) 13