ML24038A268

Enclosure 1: Request for Additional Information Round 2 (Non-proprietary)

ML24038A268
Person / Time
Site:	07109383
Issue date:	02/22/2024
From:	Storage and Transportation Licensing Branch
To:	Gesellschaft fur Nuklear-Service mbH
Shared Package
ML24038A266	List: ML24038A268
References
EPID L-2021-NEW-0002, CoC 9383, Rev 3
Download: ML24038A268 (1)

Text

Non-Proprietary Request for Additional Information Docket No. 71-9383 Model No. Castor geo69 Package Certificate of Compliance No. 9383 Revision No. 3

By letter dated July 10, 2020 (Agencywide Documents Access and Management System Accession No. ML20198M431), Gesellschaft für Nuklear-Service mbH (GNS) requested approval of its quality assurance program. By letter dated January 14, 2021 (ML21033A353), as supplemented on July 2, 2021 (ML21196A374), August 25, 2021 (ML21245A234) and March 2, 2022 (ML22073A009), GNS submitted an application for a certificate of compliance for the Model No. CASTOR geo69 spent fuel transportation package.

By letter dated March 16, 2023, staff issued a request for additional information (RAI)

(ML23046A109), to which a response was provided on August 30, 2023 (ML23257A096).

This RAI identifies information needed by the U.S. Nuclear Regulatory Commission staff (NRC or the staff) in connection with its review of the application. The staff used guidance provided in NUREG-2216, " Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, in its review of the application.

The questions below describe information needed by the staff for it to complete its review of the application and to determine whether the applicant has demonstrated compliance with regulatory requirements.

Chapter 1: General

RAI 1-1. Clarify the reason for differences in cask helium backfill pressures that are mentioned in the Part 71 safety analysis report (SAR) and Part 72 SAR operations and confirm that the packages analyses (e.g., pressure calculations, structural calculations) are bounding.

The calculations (e.g., chapter 4 pressure calculations) within the Part 71 SAR were based on the pressure listed in operation procedure SAR table 7.1-2, which indicated a cask helium backfill pressure of {proprietary information removed}.

This pressure is less than the Part 72 SAR operation procedure SAR table 9.3-1, which indicated a cask helium backfill pressure of {proprietary information removed}. A discussion that reconciled the different cask helium backfill pressures for transportation and storage was not found and, therefore, it is uncertain if an intermediate operation procedure is required for transportation, and it is uncertain if the packages analyses are bounding.

This information is needed to determine compliance with Title 10 of the Code of Federal Regulations (10 CFR) Section 71.35 and 10 CFR 71.41.

Chapter 2A: Structural Analysis

RAI 2A-1. Provide the package type for CASTOR geo69 based on the payload for which the package is designed to store.

Packages licensed in the United States (U.S.) have a package type assigned to the package design. The designer uses this package type categorization for

Enclosure 1 component design that support the package safety functions. The categorization in addition, informs the designer of the amount of radioactive material to contain, control and shield against, in maintaining the required level of package safety.

This information is required to demonstrate compliance with 10 CFR 71.33 (a)(1).

RAI 2A-2. The CASTOR geo69 transportation package SAR identifies the cask as Type B(U)-F using the requirements in 10 CFR 71.4. Explain how the cask type was translated to the design codes in Table 1.0 1 and 1.0 2.

The staff uses the guidance in NUREG/CR 6407 and NUREG/CR 3854 to ensure that the spent fuel transportation cask is designed to the level of safety as achievable using appropriate ASME codes for its design, fabrication, and inspection. The selected codes provide criteria for fabrication processes which are related to materials control, forming, heat treatment, examination, and acceptance testing. Implementation of these criteria will ensure the structural integrity of shipping containers at levels consistent with the radioactive materials being transported.

This information is required to meet the requirements of 10 CFR 71.35(a),

10 CFR 71.37(a) and 10 CFR 71.107(b)

RAI 2A-3. Use the guidance in NUREG/CR 6007 for the computation of bolt stresses or provide a comparison showing the equivalence of stress computations using VDI 2230: 2014-12 and VDI 2230: 2015-11 with bolt stress computed using NUREG/CR 6007.

The staff reviews the stress computation of highly stressed bolts using the guidance in NUREG/CR 6007 hence the bolt stress shown in the SAR will be evaluated using the methodology of the guidance document.

This information is required to meet the requirements of 10 CFR 71.51(a)(1) and (2).

RAI 2A-4. Explain how the NCT and HAC conditions envelope the condition where a canister drop from a failed lifting gear is encountered.

The lifting gear is not considered as a component important to safety (see Storage SAR section 1.2.2.1.3) and hence not addressed in the SAR. Thus, the failure of the lifting gear poses an unanalyzed condition not addressed by other safety requirements. The staff is of the opinion that this should be addressed under the handling operations that prepare the transport cask for transportation.

This information is required to meet the requirements of 10 CFR 71.

RAI 2A-5. Make the item numbering in Tab. A 2 consistent with that of SAR table 2.1-2.

Aligning the item number between the two tables allows an easy evaluation of the appropriate items of interest in the table.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

2 RAI 2A-6. Use the methodology of NUREG-6007 to compute the preload stresses in the bolts or show that the stresses computed using the VDI approach results in the same values of the stress parameters that are used in the acceptance criteria from NUREG-6007.

The staff reviews the stress computation of highly stressed bolts using the guidance in NUREG/CR 6007 hence the bolt stress shown in the SAR will be evaluated using the methodology of the guidance document.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-7. For the NCT and the HAC drop analysis, provide information on the physical properties of the foam, along each direction, used in the LS DYNA material model for the impact limiter simulations. Provide a comparison of these physical properties of the foam derived from testing to that from a simulation using the same LS-DYNA material model.

The staff needs assurance that the material models used can replicate the actual response of the foam under the specified drop conditions of NCT and HAC.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-8. For the NCT and the HAC drop analysis, provide information on how the contact area between the cask and the impact limiter was developed for the transfer of loads to the ANSYS cask and canister design models.

The impact force transferred to the cask and canister design elements from the drop is a function of the contact area post-drop. The staff needs to know how this contact area was developed from the LS-DYNA results.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-9. Provide a numeric tabulation of the deformation in the finite elements ( FEs) leading up to the contact surface between the impact limiter and the cask extending the tabulation a few FEs beyond the contact surface.

The staff wants to be sure that the drop does not contribute to an inelastic deformation of the cask body resulting in internal work done by the FEs that represent the contact boundary. This is difficult to see for the presented time stepped pictures of the foam deformation.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-10. Provide the criteria used for eroding the elements of the impact limiter as the drop proceeds.

As the drop energy is dissipated through the impact limiter some portions of the foam will be eroded i.e., unable to participate in resisting the drop. The staff

3 wants information as to what strain level was this erosion set at in the model and how does it compare with the rupture strain of the foam.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-11. Provide information on the conditions to which the aluminum cladding of the impact limiter is exposed to during contact with the rigid surface. Discuss the effect of any drag i.e., the effect of friction and the stress in the aluminum sheet and its potential to tearing. Information is requested on any increase in temperature of the aluminum sheet as a result of dragging and the state of both the aluminum and the foam when exposed to this elevated temperature.

The staff wants to know if the aluminum sheet remains intact and that the elevated temperature does not degrade the modeled performance of the foam.

This information is required to meet the requirements of 10 CFR 71.71 and 10 CFR 71.73.

RAI 2A-12. Use US equivalent codes for German design codes or show the equivalence by computing the stresses using both methods.

The staff is familiar with US codes used in the design and fabrication of spent fuel cask. While the acceptance criteria in the application uses US standards several of the design methodologies follow VDI. KTA, DIN or Eurocodes. Though these methodologies are acceptable in Europe the review staff is unfamiliar with the analysis steps used in these codes.

This information is required to ensure that subsequent compliance with the requirements of 10 CFR Part 71 assures the safety of the transportation cask.

Chapter 2B: Materials Analysis

RAI 2B-1. Provide additional information on the cast iron package coatings that demonstrate coating durability and support the emissivity values credited in the thermal analysis.

The specifications and minimum requirements for inner and outer coatings are provided in SAR section 2.2.7, which does not provide any details about materials qualification data, manufacturer data sheets, or other bases that demonstrate the performance of the coating under the elevated heat and radiation exposures of the internal package environment. The applicant states the coating is important to safety ( ITS) as the thermal analysis relies on emissivity values. However, this is not indicated in the bill of materials or the drawing package. The staff notes that the emissivity specification and how the emissivity value will be verified is important, as well as what standard the applicant is using to verify and validate these values.

This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5) and 10 CFR 71.43(d).

4 RAI 2B-2. Justify that the fuel basket plates will have adequate fracture performance in a drop accident to maintain the assumed fuel configuration in the criticality analyses.

The SAR does not include toughness testing requirements to verify that brittle fracture will not affect the structural integrity of the basket in a drop accident. The staff notes that the criticality analysis relies on the configuration of the neutron absorber plates and fuel assemblies being maintained.

Although nonferrous materials are generally excluded from fracture acceptance testing in consensus standards, the proprietary metal matrix composite is a non-code material that contains boron carbide ceramic particles that may diminish fracture performance relative to the conventional aluminum material that are considered in the ASME BPVC.

This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5)(ii), 10 CFR 71.31(c) and 10 CFR 71.55(e).

RAI 2B-3. Provide analysis to demonstrate that the package drying criteria is adequate to remove residual moisture such that it limits hydrogen generation and clarify operational procedures that would prevent accumulation of hydrogen during loading operations.

The SAR section 7.1.2 provides process steps but is unclear why this will provide adequate dryness and removal of moisture. Additionally, in t able 7.1-2 Operations for loading of contents, it is unclear if and how hydrogen levels are monitored to ensure that the flammable gas mixture is monitored and mitigated.

This information is needed to demonstrate compliance with 10 CFR 71.43(d).

RAI 2B-4. Provide additional information to demonstrate that the fuel assemblies are not exposed to a non-inert environment during the dewatering, drying and helium backfilling process during loading operations.

The SAR section 7.1.2 provides process steps, but it is not clear if this provides adequate protection of fuel assemblies without any adverse effects to the material (fuel cladding).

This information is needed to demonstrate compliance with 10 CFR 71.43(d) and 10 CFR 71.55(d)(2).

RAI 2B-5. Provide material qualification for ductile cast iron and discuss how the procurement of the cast iron cask body is controlled to ensure that the casting facility conducts required sampling and testing to qualify the material performance and ensures that the casting facility does not perform repairs of casting defects, per the ASME Boiler and Pressure Vessel (B&PV) Code criteria.

Absent such controls, describe how repairs are performed, documented, and examined to ensure mechanical properties of the cask are uniform and meet the minimum requirements of the ASME B&PV Code.

5 SAR section 1.2.1.1 states that the cask body is made of ductile cast iron, composed of a hollow cylinder with a closed bottom end, cast in one piece.

ASME B&PV Code, S ection III, Division 3, WB-2573 states the following:

Casting shall not be repaired by plugging, welding, brazing, impregnation, or any other means.

The staff notes that these ASME B&PV Code requirements are in place to eliminate specific issues with each of the repair methods to potentially degrade the microstructure and mechanical properties of ductile cast iron. If controls are not in place to ensure that defects are not repaired, the staff requests information on how defects will be repaired and examined to ensure that code-required properties will be achieved in these localized areas.

The information is needed to determine compliance with the regulatory requirements in 10 CFR 71.31(c).

Chapter 3: Thermal Analysis

RAI 3-1. Discuss the thermal impact of package temperatures and pressures if the surface enhancement factor was applied to the radiant energy component during the 30- minute engulfing fire.

SAR section 3.3.1.4 indicated that the packages radial fins were not explicitly modeled; rather, a surface enhancement factor was applied to the heat transfer coefficient at the models corresponding unfinned surface. In addition, SAR figure 3.4-1 indicated that radiant heat transfer is the dominant heat flux component to the package. However, there was no discussion regarding the impact of increased thermal input to the fins additional area from increased radiation heat transfer during the fully engulfing fire accident condition. Staff notes that the concept of angle of incidence, as indicated in RAI response 3-11 and mentioned in SAR s ection 3.3.1.4, generally would not be a factor for the assumed fully engulfing fire.

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-2. Clarify the process for determining the need for a personnel barrier if thermal measurements during pre-shipment operations indicate excessive surface temperatures and confirm the personnel barrier would not adversely impact package thermal performance.

a. While reviewing the responses to RAIs, staff noted that SAR table 7.1-3 (footnote 2) was revised; it stated that a personnel barrier is to be placed on the package if measurements during pre-shipment operations indicate surface temperatures (or doses) do not meet regulations (e.g., 10 CFR 71.43, 10 CFR 71.47). The procedure did not provide instruction to first investigate the reason for the excessive temperature (e.g., incorrect loading, source term

6 methodology1) in order to confirm the package is in compliance with the certificate of compliance.

b. SAR section 3.3.2.3 indicated that thermal model results showed margin between the maximum package accessible surface and the 85 deg C exclusive use shipment requirement of 10 CFR 71.43(g). Likewise, SAR section 3.3.1.5.4 and section 3.3.2.1 listed conservative assumptions which should indicate the maximum package accessible surface temperature is conservative. However, the SAR did not offer potential explanations for a surface temperature measurement exceeding the 85 deg C exclusive use temperature limit.

c. There was no thermal analysis to demonstrate acceptable package thermal performance with a personnel barrier, recognizing that heat transfer from the package could be adversely impacted depending on the personnel barrier design.

This information is needed to determine compliance with 10 CFR 71.35(a),

71.43(g), 71.47.

RAI 3-3. Provide the comparison inputs and calculation that demonstrate CASTOR geo69 pressures are bounded by the assumed 0.15 fission gas release fraction of high-burnup fuel, relative to low-burnup fuel that has a 0.30 percent fission gas release fraction.

The response to RAI 3-4 and SAR section 4.2.1 appeared to indicate that the high-burnup fuels listed in SAR table 1.2-12 result in greater fission gas amounts and canister pressures than low burnup fuels. However, there was no calculation for staff to review and make findings.

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-4. Clarify in SAR chapter 7 Operations that the condition and cleanliness of the fins are maintained prior to shipment to ensure fin effectiveness during transport.

The SAR section 3.3.1.4 indicated that the effectiveness of the cask fins is an important factor to achieve thermal performance of the package. Fin effectiveness is dependent on maintaining fin geometry (i.e., no large areas of deformed fins) and cleanliness (i.e., no fouling factors, maintaining emissivity factors). However, the response to RAI 3-2 did not provide explicit instruction in SAR table 7.1-3 to repair the fins if the H3 inspection step finds unacceptable fin condition (e.g., deformations) and cleanliness (e.g., fouling factors).

This information is needed to determine compliance with 10 CFR 71.35(a).

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Reference:

R. Cumberland et. al., A Study on the Relationship between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks, Oak Ridge National Laboratory, June 17, 2020.

7 RAI 3-5. Clarify in the SAR the basis for applying approximately 18.4 kW decay heats (rather than the 18.5 kW and approximate {proprietary information removed}

bounding decay heat described in SAR section 3.1.2 and section 3.3.2.1) on the thermal models.

SAR sections 3.3.2.1, 3.4, 3.5, and 3.6 indicated that thermal analyses considered decay heats less than 18.5 kW. However, t he basis for demonstrating a loading maximum decay heat of 18.5 kW, as noted in SAR section 3.1.2, was not clearly described or shown in the results.

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-6. Clarify that the effective thermal conductivity applied to the 3-dimensional (3D )

model baskets homogenized fuel correctly considers the appropriate amount of thermal radiation heat transfer.

The response to RAI 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. These thermal analyses, including those for normal conditions of transport and hypothetical accident conditions, were based on SAR section 3.3.1.5.2, which stated that radiation heat transfer among fuel rods and inner surfaces of the fuel channel, and among the fuel channel and the inner surface of the basket sheets is accounted for in the detailed 2-dimensional (2D )

model and the simplified 2D model. However, SAR figure 3.3-7 indicated that the half-symmetric 3D model of the canister and cask includes the fuel channel, the helium between the fuel rods and fuel channels, and the helium between the fuel channels and outer basket sheets (i.e., as part of the homogenized fuels effective thermal conductivity and explicitly modeled in the 3D ANSYS model).

This may appear to indicate that radiation heat transfer within the basket is accounted for twice, which would not be an accurate or bounding assumption.

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-7. Demonstrate as part of the SAR chapter 3 thermal analyses that spent fuel baskets and other components that are important to safety (i.e., canister, transfer cask, cask loading unit) for operations within the reactor building would not be affected by the high temperature listed in SAR section 3.6.4.3 and without the use of air conditioning within the reactor building. In addition, clarify if the following considerations would impact the conclusions in SAR section 3.6.6 and section 3.6.7 regarding there being no temporal restrictions and no need for additional calculations for transfer operations inside the reactor building and loading of the canister in the cask(s).

a. The response to RAI 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. Staff notes that Part 72 SAR section 12.1.2 stated that off-normal temperatures during cask loading unit (CLU) handling operations in the reactor building are not credible and Part 72 SAR section 2.2.5.1 indicated there are no off-normal environmental temperatures within the reactor building because it is assumed that the building is air conditioned. However, active cooling of the CASTOR geo69 systems heat sink (i.e., internal reactor building temperature) via air conditioning is inconsistent with regulations requiring only passive cooling.

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b. Part 72 SAR section 12.1.2 stated that it was assumed that off-normal temperatures during CLU handling operations in the reactor building are covered by normal temperature evaluations for the Part 72 application.

Although the thermal analysis of the off-normal storage condition at a

{proprietary information removed} ambient temperature discussed in Part 72 SAR section 4.5.4 indicated that ITS components were below allowable temperatures, it was not demonstrated that the content and ITS component temperatures within the canister and transfer cask (which does not include fins) is bounded by the finned CASTOR geo69 storage system thermal analysis at the potential higher 52 deg C ambient temperature. Likewise, the

{proprietary information removed} to begin the dewatering process discussed in the Part 71 SAR section 3.6.3 would be reduced if {proprietary information removed} (per item a, above).

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-8. Clarify and demonstrate as part of your thermal analysis in SAR chapter 3 the appropriateness of the thermal model assuming no gaps between transfer cask components.

The response to RAI 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. Although Part 72 SAR section 1.2.2.1.1 indicated the transfer cask is fabricated from a number of radial sections (e.g., outer shell, water jackets, lead), Part 72 SAR section 1.2.2.1.6 indicated an absence of air gaps in the transfer cask body. Likewise, a visual review of the ANSYS model indicated there were no contact resistances between the radial components.

However, there was no discussion that demonstrated assurance that fabrication would be possible without the presence of gaps between components (e.g., inner liner and lead shielding, which is poured or composed of multiple cast parts or lead sheets, per Part 72 SAR section 10.1.6.3) and the corresponding thermal contact resistance that would result in increased component temperatures. For example, Part 71 SAR table 3.6-5 appears to indicate that increased thermal resistance through the transfer cask (i.e., presence of gaps within transfer cask) may result in some ITS components (e.g., basket sheets, polyethylene shielding CLU components) being above allowable temperatures.

This information is needed to determine compliance with 10 CFR 71.35(a).

RAI 3-9. Demonstrate and clarify in the short -term section 3.6 of the SAR that all ITS components are within allowable values and whether there are time limits during those operations, including when the canister is within the transfer cask on top of the CASTOR geo69 cask using the cask loading unit.

The revised SAR section 3.6.7 of the recent submittal indicated that there was no need to analyze the CASTOR geo69 system with the cask loading unit, such as the condition of the transfer cask containing the fuel rod filled canister being positioned on top of the CASTOR geo69 cask with the bottom lid closed (as described in SAR section 1.2.4.3.4) and in which the canister would be enclosed in the transfer cask (which has no cooling fins and no improved external convection heat transfer capability). However, there were no temperature comparisons of the ITS components with their allowable temperatures

9 (e.g., polyethylene shielding material within the transfer lock, per Part 72 SAR section 1.2.2.1.2) and no discussion of potential time limits.

This information is needed to determine compliance with 10 CFR 71.35(a).

Chapter 4: Containment Analysis

RAI 4-1. Revise Table 1.0-2 List of BPVC Alternatives for the CASTOR geo69 DSS Transport Cask on page 1.0 -9 and provide additional justification for the request for the approval of a significant departure from the ASME B&PV Code Section III, Division 3, WB-6120.

Table 1.0- 2 List of BPVC Alternatives for the CASTOR geo69 DSS Transport Cask on page 1.0 -9, provides a justification for an alternative to ASME B&PV Code Section III, Division 3, WC-6120, for hydrostatically, or pneumatically pressure testing of the transport system. The justification provided includes the following statement:

{proprietary information removed}

The requirements of ASME B&PV Code Section III, Division 3, WB-6120, applicable to transportation containments, include pressure testing in accordance with WB-6200 (hydrostatic test) or WB-6300 (pneumatic test) and then, in addition, helium leak testing in accordance with WB-6700. The alternative listed in table 1.0-2 does not address the combined requirement of pressure testing and helium leak testing required for a transportation containment designed, constructed, and tested in accordance with the requirements of ASME B&PV Code Section III, Division 3, Subsection WB Class TC Transportation Containments. This exception, if granted by the NRC as requested, would provide relief from the ASME code requirement for any pressure testing of the geo69 transportation containments in the course of serial production of both the cask and canister, based solely on the structural analysis of the geo69 transport cask and canister presented in section 2 of the SAR. The alternative proposed represents a significant departure from the ASME code requirements, one that NRC staff has not previously granted to any applicant for certification of a transportation package design; therefore, the alternative requested would need additional justification in order to be further considered by the staff.

This information is needed to determine compliance with 10 CFR 71.31(c) and 71.85(a).

Chapter 17: Quality Assurance Program Desc ription Analysis

The following questions are on the GNS -Quality Assurance Program Description (QAPD) submitted August 30, 2023 (ML23257A096).

RAI 17-1. Revise chapter 17 to clarify and describe what applicable codes and standards are used in the storage, retention, and maintenance of both electronic and hardcopy quality assurance records.

The GNS QAPD chapter 17 states, in part, the requirements and responsibilities for records transmittal, retention (such as duration, location, fire protection, and

10 assigned responsibilities), and maintenance shall be consistent with applicable codes, standards, and procurement documents. The staff needs to know what the applicable codes and standards are and if this information is within the procedures and instructions that control the storage, retention, and maintenance for both electronic and hardcopy quality assurance records. The NRC regulatory guide (RG) 7.10, revision 3, Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material, and RG 1.28, revision 4, Quality Assurance Program Criteria (Design and Construction), which endorsed Nuclear Quality Assurance (NQA)-1-2008, and NQA-1a-2009 Addenda provide regulatory information about the control of electronic records. Both RGs provide reference to an NRC, Regulatory Issue Summary 00- 18, Guidance on Managing Quality Assurance Records in Electronic Media, (ML003739359), which provides applicants and licensees with a way to satisfy the requirements for the maintenance of quality assurance records.

This information is needed to demonstrate compliance with 10 CFR 71.135.

Revise chapter 18.2 to remove a description that allow the internal audit interval to be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities (i.e., the quality assurance program (QAP) elements) are being satisfactorily accomplished. The NRC RG 7.10, revision 3, and RG 1.28, revision 4, that endorse the ASME NQA-1 took exception to the statement that the interval may be extended up to two (2) years.

The GNS QAPD chapter 18.2 states, in part, that all applicable QAP elements shall be audited at least once each year or at least once during the life of the activity, whichever is shorter. Further, section 18.2 goes on to state that this interval may be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities are being satisfactorily accomplished in accordance with the applicable QAP elements. This language is consisted with the ASME NQA-1-2008 and NQA-1a-2009 Addenda, Requirement

18. However, in NRC RG 7.10, revision 3, Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material, and RG 1.28, revision 4, Quality Assurance Program Criteria (Design and Construction),

which endorsed NQA-1-2008, and NQA-1a-2009 Addenda, Requirement 18, provide a regulatory position that all applicable QAP elements shall be audited at least once each year or at least once during the life of the activity, whichever is shorter. The NRC RGs took exception to the statement that the interval may be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities are being satisfactorily accomplished.

This information is needed to demonstrate compliance with 10 CFR 71.137.

Revise chapter 18.3.2 to provide a description on how the lead auditor qualifications meet the NRC RG exception provided in RG 1.28, revision 5 as committed to and described in ASME NQA-1-2015. For example, the description can be as simple as the lead auditor qualifications and certifications are described in implementing procedures and account for the exceptions identified in NRC RGs for endorsed ASME NQA-1 standards.

The RG 1.28, revision 5 endorses ASME NQA-1 2015 with certain clarifications and exceptions. The exception to the lead auditor qualifications is as follows:

11 Prospective lead auditors, with comparable industry experience

[emphasis added], may satisfy the lead auditor qualification requirement of participating in a minimum of five quality assurance audits within a period of 3 years prior to the date of qualification by alternatively demonstrating the ability to properly implement the audit process, effectively organize and report results, and participate in at least one nuclear audit within the year preceding the date of qualification, subject to review and acceptance by the responsible QA organization [emphasis added].

This information is needed to demonstrate compliance with 10 CFR 71.105 and 10 CFR 71.137.

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