ML23020A921

Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks Request for Additional Information Non-Proprietary Responses

ML23020A921
Person / Time
Site:	07201004
Issue date:	01/20/2023
From:	Narayanan P Orano TN Americas
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML23020A920	List: ML23020A921 ML23069A285 ML23069A286 ML23069A287 ML23069A288
References
EPID L-2022-LLA-0079, E-61864, CAC 001028
Download: ML23020A921 (1)

Text

January 20, 2023 E-61864 Orano TN 7160 Riverwood Drive U. S. Nuclear Regulatory Commission Suite 200 Attn: Document Control Desk Columbia, MD 21046 USA One White Flint North Tel: 410-910-6900 11555 Rockville Pike Fax: 434-260-8480 Rockville, MD 20852

Subject:

Response to Request for Additional Information - Application for Amendment 18 to Standardized NUHOMS Certificate of Compliance No.

1004 for Spent Fuel Storage Casks, Revision 2 (Docket No. 72-1004, CAC No. 001028, EPID: L-2022-LLA-0079)

Reference:

[1] Letter from Chris Allen (NRC) to Prakash Narayanan (TN Americas LLC), Review of Amendment No. 18 to Certificate of Compliance No. 1004 for the Standardized NUHOMS System - Request for Additional Information (EPID: L-2022-LLA-0079), dated December 21, 2022

[2] Letter E-60447 from Prakash Narayanan, Application for Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 0 (Docket No. 72-1004), dated May 20, 2022 TN Americas LLC (TN) hereby submits our response to the Request for Additional Information (RAI) forwarded by Reference [1]. Enclosure 2, herein, provides a proprietary version of the responses to the RAIs with the exception of RAI 4.1. The response to RAI 4.1 will be provided during the week of January 23, 2023. Enclosure 3 provides a public version of the responses to the RAIs.

Enclosure 4 provides a complete revision of the CoC Appendix B Technical Specifications (TS) with proposed changes tracked. The proposed proprietary version of the CoC 1004 Amendment 18, Revision 2 changes to the Standardized NUHOMS System Updated Final Safety Analysis Report (UFSAR) changed pages are included as Enclosure 5, with a footer on each page annotated as 72-1004 Amendment 18, Revision 2, January 2023, and changes indicated by italicized text and revision bars. The changes are further annotated with gray shading or a gray box enclosing an added section, as well as a footer to distinguish the Amendment 18, Revision 2 changes from previous Amendment 18 changes.

Revised drawings show alpha-numeric revision numbers and clouds surrounding the changed drawing information. The public version of these UFSAR changed pages and drawings is provided as Enclosure 6.

Enclosure 7 provides a new scope item related to use of blended concrete in accordance with ASTM C595 for HSM fabrication and is not associated with the RAI responses. UFSAR changes associated with blended concrete will also be indicated by italicized text and revision bars and will be annotated as being associated with the additional scope item.

Enclosures transmitted herein contain SUNSI. When separated from enclosures, this transmittal document is decontrolled.

E-61864 Document Control Desk Page 2 of 2 also describes two minor editorial corrections that have been made to the TS. Enclosure 8 provides an updated evaluation of impacts on CoC 1004 renewal, which was previously provided in Reference [2], to address blended concrete.

Certain portions of this submittal include proprietary information, which may not be used for any purpose other than to support the NRC staffs review of the application. In accordance with 10 CFR 2.390, TN Americas LLC is providing an affidavit (Enclosure 1), specifically requesting that this proprietary information be withheld from public disclosure.

Should you have any questions regarding this submittal, please do not hesitate to contact Mr. Douglas Yates at 434-832-3101 or me at 410-910-6859.

Sincerely, Prakash Narayanan Chief Technical Officer cc: Chris Allen, NRC DFM

Enclosures:

1. Affidavit Pursuant to 10 CFR 2.390

2. RAI and Responses (Proprietary)

3. RAI and Responses (Public)

4. Proposed CoC Appendix B Technical Specifications, CoC 1004 Amendment 18, Revision 2

5. Proposed Amendment 18, Revision 2 Changes to the Standardized NUHOMS System Updated Final Safety Analysis Report (Proprietary)

6. Proposed Amendment 18, Revision 2 Changes to the Standardized NUHOMS System Updated Final Safety Analysis Report (Public)

7. Description of Amendment 18 Changes Not Related to RAIs

8. Impacts of New Blended Cement Scope Item on CoC 1004 Renewal

Enclosure l to E-61864 AFFIDAVIT PURSUANT -

TO 10 CFR 2.390 TN Americas LLC )

State of Mary land) ss.

County of Howard )

I, Prakash Narayanan, depose and say that I am the Chief Technical Officer of TN Americas LLC, duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information that is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR 2.390 of the Commission' s regulations for withholding this information.

The information for which proprietary treatment is sought meets the provisions of paragraph (a) (4) of Section 2.390 of the Commission's regulations. The information is listed below:

• Enclosure 2 - Portions of Proposed Amendment 18, Revision 2 Responses to RAis *

• Enclosure 5 - Portions of Proposed Changes to the Standardized NUHOMS System Updated Final Safety Analysis Report This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

1) The information sought to be withheld from public disclosure involves portions of the RAJ Responses and the UFSAR, all related to the design of the Standardized NUHOMS dry spent fuel storage system, which are owned and have been held in confidence by TN Americas LLC.

2) The information is of a type customarily held in confidence by TN Americas LLC, and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.

3) Public disclosure of the information is likely to cause substantial harm to the competitive position of TN Americas LLC, because the information consists of descriptions of the design of dry spent fuel storage systems, the application of which provide a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their product' s position or impair the position of TN Americas LLC's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

Further the deponent sayeth not.

Prakash Narayanan Chief Technical Officer, TN Americas LLC Subscribed and sworn before me this I~~ y of January, 2023 .

~

Notary Public KHVNESVA TAYLOR Notary Public Howard County My Commission Expires lD__;__S_;~ Maryland My Commission Expires Oct. 5, 2025 Page 1 of 1

Enclosure 2 to E-61864 RAI and Responses (Proprietary)

Withheld Pursuant to 10 CFR 2.390

RAIs and Responses - Public Enclosure 3 to E-61864 Structural:

RAI 4.1:

Provide complete licensing drawings for the NUHOMS 24PTH dry shielded canister (DSC) Type 3 basket.

The drawings provided in section P.1.5 (reference 1) do not contain the necessary dimensional information on key components of the basket (e.g., R90 transition rails, R45 transition rails, R45 angle plates, steel plates, aluminum plates, and metal matrix composite plates) to evaluate the geometry, interlocking features, and numerical models. Reference dimensions alone are insufficient. The complete licensing drawings, including the missing dimensional information (including nominal dimensions and associated tolerances on thicknesses of components and overall basket and rail lengths) for the 24PTH DSC basket assemblies, are needed for the staff to completely evaluate the structural, criticality, and shielding performance impacts.

The above information is necessary to comply with Title 10 of the Code of Federal Regulations (10 CFR) 72.230(a); 72.124(a) and (b); and 72.236(b), (c) and (d).

Response to RAI 4.1:

The response to this RAI item will be submitted at a later date.

Page 1 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Proprietary Information on Pages 2 through 4 Withheld Pursuant to 10 CFR 2.390 Page 2 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 4.3:

Provide justification for not using a dynamic load factor (DLF) for ANSYS quasi-static analysis of the basket cross-section model.

Quasi-static analyses use a DLF to capture the added deformation of dynamically applied loads as compared to statically applied loads (reference 3). The hypothetical accident side drop was evaluated with the quasi-static ANSYS cross-section model for a 75g maximum acceleration, but no DLF was used in the models or discussed in the application. The use of a DLF accounts for uncertainties involved in natural frequency, load duration, and load time history shape which depend on the physical characteristics of the fuels, the basket, the containment, and the cask.

However, the applicant did not use a DLF in the analysis, which implies that there are no dynamic effects on a fuel system in a cask during a drop.

Provide a justification for not using a DLF for the quasi-static analysis approach. Correct evaluation of basket behavior (e.g., deformations) is important for and impacts, among others, the criticality and shielding analyses.

The above information is necessary to comply with 10 CFR 72.236(c) and (d).

Response to RAI 4.3:

As discussed in Section 3.2.2.3 of CoC 1004 SER [1], the DLFs for the vertical, horizontal, and corner drops of the cask are 1.50, 1.75, and 1.25, respectively. The resulting deceleration values obtained by multiplying the unfactored levels by these factors are 73.5g, 66.5g, and 25.0g for the three drop orientations. As discussed in Section P.3.7.4.1 of the UFSAR, for the horizontal side drop, the quasi-static system response to the conservative deceleration load of 75g is considered to bound any dynamic effects including the use of a DLF.

Reference:

1. Safety Evaluation Report of Vectra Technologies, Inc., a.k.a. Pacific Nuclear Fuel Services, Inc. Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, U.S. Nuclear Regulatory Commission, December 1994.

Impact:

No change as a result of this RAI.

Page 5 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 4.4:

Provide buckling evaluations of the 24PTH DSC and Type 3 Basket.

Buckling of the 24PTH DSC and the Type 3 basket were not evaluated. NUREG-2215 suggests that buckling of the fuel basket materials should be considered for normal and off-normal loading (section 4.5.5.1.2 of reference 2) and buckling of the canister should be considered for a load drop (section 4.5.6.1.1 of reference 2). Buckling evaluations of the 24PTH DSC and Type 3 Basket are needed for the staff to complete the evaluations. The resulting basket deformation from these evaluations should be appropriately included in the criticality analysis and shielding analysis since the basket affects axial dose rates.

The above information is necessary to comply with 10 CFR 72.236(c) and (d).

Response to RAI 4.4:

For the 24PTH DSC Shell, the buckling evaluations of the 24PTH DSC for a postulated 60g end drop are provided in UFSAR Section P.3.7.4.2.4. This section remains unchanged in this submittal since the same DSC Shell is used for all basket types of the 24PTH DSC.

For normal and off-normal loading, the stress in the Type 3 basket compartment plates for the 1g axial handling load due to conservative basket and fuel assembly weights is 0.54 ksi. This stress value is low, such that the DW + 1g axial handling load combination does not control, and no further buckling evaluation is required.

The buckling evaluation for the Type 3 basket under accident side drop loading is discussed in UFSAR Section P.3.8.7.3, and results are presented in UFSAR Tables P.3.8-8 and P.3.8-9. The buckling evaluation for the Type 3 basket under end drop loading is provided in UFSAR Section P.3.8.7.4. Both of these evaluations demonstrate that the 24PTH Type 3 basket will not buckle under accident conditions.

Impact:

No change as a result of this RAI.

Page 6 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 4.5:

Provide justification that the cross-section model approach is appropriate for the 24PTH DSC Type 3 basket.

The interlocking egg-crate design of the Type 3 basket utilizes no welding and relies on attachments at the basket exterior to keep the structure intact. The mechanical interlocking of the plates at the center of the basket likely has clearance gaps for assembly and thermal growth that give the basket additional flexibility (drawing details were not available to evaluate the gap dimensions). The application has little discussion of the joints and gaps between the basket plate structures and how these are conservatively accounted for in the ANSYS cross-section model. It is not apparent that the short cross-section basket model accounts for the flexibility of the interlocking joints in the series of plates forming the egg-crate design, and the boundary conditions may be artificially stiffening the structure by disallowing any axial motion of the plates under flexure that could occur due to axial clearance gaps. If the cross-section model approach is overly stiff for simulation of an egg-crate design, this could lead to underprediction of the basket compartment relative displacements, which is important for and impacts, among others, the criticality and shielding analyses.

Provide justification that the modeling domain with the cross-section approach appropriately captures the inherent joint flexibility and assembly gaps of the Type 3 basket design to predict maximum compartment displacement. In addition, provide justification for not using the LS-DYNA approach as done with the Type 1 and Type 2 basket design.

The above information is necessary to comply with 10 CFR 72.236(c) and (d).

References:

1. Orano TN, Application for Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 0 (Docket No. 72-1004), May 20, 2022.

2. NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities -

Final Report, April 2020.

3. NUREG/CR-3966, Methods for Impact Analysis of Shipping Containers, November 1987.

Response to RAI 4.5:

The mechanical interlocking of the plates at the center of the basket has clearance gaps for assembly and thermal growth, which are considered in the 6-inch cross-section basket model using [ ] elements as described in UFSAR Section P.3.8.6.1. [

] The key basket dimensions are provided in Drawings NUH24PTH-S-5012-SAR, NUH24PTH-L-5012-SAR, and NUH24PTH-S-LC-5012-SAR in UFSAR Section P.1.5, for the 24PTH-S, -L, and -S-LC DSC, respectively. The drawings show the following basket plate thicknesses.

Page 7 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 Page 8 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 The 24PTH Type 1 and Type 2 basket designs are welded structures for which LS-DYNA was used for structural analysis. However, the 24PTH DSC Type 3 basket egg-crate design is based on the NUHOMS EOS System baskets, so the analysis methodologies including the use of ANSYS for structural analysis for the 24PTH DSC Type 3 basket followed the approach for the NUHOMS EOS basket analysis described in CoC 1042 UFSAR, Section 3.9.2 [1].

Reference:

1. NUHOMS EOS System Updated Final Safety Analysis Report, Docket Number 72-1042, Revision 4.

Impact:

No change as a result of this RAI.

Page 9 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Thermal:

RAI 5.1:

Clarify the following portion of section 4.3.6 b in appendix B of the Standardized NUHOMS CoC No. 1004 Technical Specifications (TS), (3) a comparison of the inlet and outlet temperature difference to predicted temperature differences for each individual horizontal storage module (HSM) or HSM-H.

The staff notes, there is a table on page 10 of the Standardized NUHOMS CoC No. 1004 appendix A TS, section 4.4, that describes the maximum air temperature rise through the HSM that is allowed as a function of the decay heat load of the DSC and the HSM model. If that table were removed, as proposed in proposed change #2, there would no longer be an inlet and outlet maximum air temperature rise criteria that is needed for (3) in the Standardized NUHOMS CoC No. 1004 appendix B TS section 4.3.6 b to assure a positive means to identify conditions which threaten to approach temperature criteria for proper HSM or HSM-H operation and allow for the correction of off-normal thermal conditions that could lead to exceeding the concrete and fuel clad temperature criteria.

This information is needed to assess compliance with 10 CFR 72.236(f).

Response to RAI 5.1:

Although the table identifying maximum air temperature rise through the HSM based on maximum decay heat load was removed as part of the deletion of Inspections, Tests and Evaluations (ITE) Appendix A, Section 4.4, the table will be added to Appendix B TS Section 4.3.6.b to provide the maximum air temperature rise criteria in the relevant TS. In addition, references to ITE Section 4.4 have been deleted from UFSAR pages 10-11, T.8-13, U.8-12, W.8-25, Y.8-13, and Z.8-12a.

Impact:

Technical Specifications Section 4.3.6 has been revised as described in the response.

UFSAR Sections 10, T.8.1.6, U.8.1.6, W.8.1.6, Y.8.1.6 and Z.8.1.6 have been revised as described in the response.

Page 10 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Shielding:

RAI 6.1:

Provide shielding integrity tests, including acceptance criteria, for the lead in the DSC shield plugs, and ensure the shielding analysis accounts for the impacts of the acceptance criteria on shielding performance and dose rates.

Section P.9.1.5 of the UFSAR includes the shielding integrity tests for the NUHOMS system with the 24PTH DSC. It states that the DSC shield plugs are steel. While two variants have steel shield plugs (the 24PTH-S and 24PTH-L), the 24PTH-S-LC has lead shield plugs. The lead may either be poured lead or precast and machined to fit into the shield plug cavity. Thus, the shielding integrity tests need to be modified to address the lead used in the 24PTH-S-LC DSCs shield plugs, using appropriate test methods and acceptance criteria. The applicant should also ensure that the shielding analysis appropriately addresses the impacts of these criteria for normal operations, off-normal conditions, and accident conditions dose rates and doses.

This information is needed to demonstrate compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.1:

A clarification has been made to the existing wording in Section P.9.1.5 of the UFSAR that states The gamma and neutron shielding materials of the storage system itself are limited to concrete HSM components and steel shield plugs in the DSC to now indicate that this does not account for the 24PTH-S-LC DSC. Further clarification is made that the lead pour or precast insert will undergo a volumetric inspection to verify that the effective thickness of the lead through any section conforms to the thickness specified on the drawings. This is to be done using a gamma inspection or an ultrasonic inspection. Since the lead is being inspected and verified at all sections, there is no impact of these criteria on the shielding analysis. Additional clarification has been made to Section P.5.4.7.4 of the UFSAR to indicate that 3% reduction in the lead density is employed in the shielding model to account for potential voids in the lead.

Impact:

UFSAR Sections P.5.4.7.4 and P.9.1.5 has been revised as described in the response.

Page 11 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.2:

Provide further justification for the proposed change to CoC appendix B, section 4.3.2 to remove verification and monitoring of the transfer casks liquid neutron shield for draining operations of the DSC cavity. Also, modify the text of the condition as described below.

The condition to verify the transfer casks liquid neutron shield is filled and to monitor it during draining operations of the DSC cavity and the DSC-TC annulus were introduced in Amendment 11 for the NUHOMS system (see safety evaluation report (SER) section 6.5 (ADAMS Accession No. ML14010A486) for Amendment 11, rev 0.). This resulted from an instance where the neutron shield was inadvertently drained during cask operations. The applicants only justification for the currently proposed change is that the DSC cavity drain is on the opposite end of the DSC from the annulus and neutron shield drains. Based on the staffs review of the transfer cask drawings, the validity of the justification is not clear since the drawings indicate valves on the neutron shield that are on the same end of the DSC as the DSC cavity drain. Its also not clear how the current justification is sufficient to address the staffs review that led to this requirement as documented in the Amendment 11 SER. Further, there should be other means in addition to location differences to ensure against accidental draining of the neutron shield when the intention is to drain the DSC cavity. These means should include items such as operating procedures, sufficient descriptions of which are included in the UFSAR, with justification provided as to why these means are sufficient to ensure that even with the proposed technical specification change, inadvertent draining of the neutron shield will not occur.

Also, the text regarding this monitoring under the heading TCs with a Liquid Neutron Shield, Other Than the OS197L TC should be modified to read as: When draining , verify the NS is filled when these draining operations are initiated and monitor the NS continuously (italics indicate requested text changes) The required actions should be consistently described with the description given in the last paragraph under the heading OS197L TC. Additionally, ensure the acronym NS is defined on the first use.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d) and ensure the system is sufficient to facilitate licensee compliance with 10 CFR 20.1101 and 20.1201(a) and 10 CFR 72.126(a)(6).

Response to RAI 6.2:

TN requested this change based on the dry shielded canister (DSC) cavity not having a bottom drain or any other bottom connection. The DSC is not drained, but is either pumped down or the water is removed by forcing helium into the DSC and pushing the water out. Those evolutions utilize fixtures, etc., at the top of the DSC. Since the liquid neutron shield (NS) drain is at the bottom of the neutron shield tank, this change appeared to be reasonable .

However, because the technical specifications (TS) and the updated final safety analysis report (UFSAR) use the terms drain, draining, and draining operations for (1) the NS, (2) the transfer cask (TC)/DSC annulus, and (3) the DSC cavity, there is the possibility of human error in the proper draining of the NS, the TC/DSC annulus, and the DSC cavity. Based on this, and the importance of radiation safety and the shielding provided by the NS, TN is withdrawing the request to delete DSC cavity from TS 4.3.2. The associated revised text has been reverted to the Amendment 17 language.

Page 12 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Regarding the changed language requested in the RAI, for consistency between the split-out requirements that apply to the OS197L TC, versus other TCs with liquid neutron shields, the text revisions requested have been made.

Impact:

Technical Specifications Section 4.3.2 has been revised as described in the response.

Page 13 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.3:

Provide further justification for the proposed change to the first paragraph of CoC appendix B, section 4.3.2 to remove the reference to 10 CFR Part 20.

The applicant is proposing to remove the requirement that the licensee, as part of its 10 CFR 72.212 evaluation, perform an analysis to confirm the limits of 10 CFR Part 20 will be satisfied.

The applicant states that this is because the purpose of the 72.212 evaluation is to confirm compliance with 10 CFR 72.104 limits. While that is the purpose in the regulation, the CoC and its appendices can specify that the evaluation also address other items as needed. This particular requirement was added as part of Amendment 11 to the NUHOMS system. A significant aspect of that amendment was the addition of the OS197L transfer cask, for which there were significant shielding and radiation protection concerns, as described and evaluated in the SER for that amendment (ADAMS Accession No. ML14010A486). In reformatting section 4.3.2, it appears some of the significance and purpose of that requirement (i.e., confirming limits of Part 20 will be satisfied as part of the 10 CFR 72.212 evaluation) is no longer clear. Given the concerns with the transfer casks in the NUHOMS system, particularly the OS197L, that led to the inclusion of this and other requirements, additional justification is needed for removing this requirement. That justification should address those related points that the staffs requests for additional information (see ML080020420, ML080020446, ML110320385, ML13149A438, ML092330146, ML102230097, ML102230099, and ML110590060 for the RAI responses, some of which are proprietary; the RAIs are included in these responses) and the SER described in the Amendment 11 review.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d) and ensure the system is sufficient to facilitate licensee compliance with 10 CFR 20.1101 and 20.1201(a) and 10 CFR 72.126(a)(6).

Response to RAI 6.3:

After careful consideration of the RAI responses to Amendment 11 and the Amendment 11 SER, as referenced in the above request for additional information (RAI), TN has withdrawn the proposed change to remove the reference made to 10 CFR Part 20 in CoC Appendix B, Section 4.3.2. The associated TS 4.3.2 revised text has been reverted to the Amendment 17 language.

Impact:

Technical Specifications Section 4.3.2 has been revised as described in the response.

Page 14 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.4:

Justify the modification of the dose rate limits for the transfer casks and the storage modules to combine the 24PTH-S-LC with the 24PTH-S and 24PTH-L in the transfer cask and in the HSM-H.

It is not clear that the dose rate limits for the 24PTH-S and 24PTH-L DSCs for both the transfer cask and the HSM-H are also appropriate limits for the 24PTH-S-LC. Given the differences in the DSCs and the HSM-H versus the Standardized HSM, appropriate limits for the 24PTH-S-LC would be significantly less than for the 24PTH-S and 24PTH-L, by a factor of three on the radial transfer cask surface and about a factor of ten on the HSM-H front bird screens. Thus, the proposed changes would result in non-conservative dose rate limits for the 24PTH-S-LC DSC in the transfer cask and in the HSM-H storage module. The staff further notes that the dose rate limits for the 24PTH-S-LC DSC have been kept separate from the limits for the 24PTH-S and 24PTH-L DSCs from the time the 24PTH DSC was added to the Standardized NUHOMS system. So, it is not clear what has changed to support combining the limits for the transfer cask and using the limits for the 24PTH-S and 24PTH-L in the HSM-H for the 24PTH-S-LC in the HSM-H.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d) and ensure the system is sufficient to facilitate licensee compliance with 10 CFR 20.1101 and 20.1201(a) and 10 CFR 72.126(a)(6).

Response to RAI 6.4:

Dose rates limits for the 24PTH system are reported in CoC 1004 Appendix A Inspections, Tests, and Evaluations (ITE) Section 3.2 for the transfer cask (TC) and in Section 3.3.2 for the Standardized HSM and HSM-H.

For the 24PTH-S-LC Dry Shielded Canister (DSC) in the Standardized Horizontal Storage Module (HSM), the dose rate limits at the three locations defined in CoC 1004 Appendix A ITE Section 3.3.1 are:

• 600 mrem/hr at HSM front bird screen

• 105 mrem/hr at outside door and

• 400 mrem/hr at end shield wall exterior while for the 24PTH-S and 24PTH-L DSC as 24PTH DSC in HSM-H, the dose rate limits at the three locations are:

• 1400 mrem/hr at HSM front bird screen

• 5 mrem/hr at outside door and

• 20 mrem/hr at end shield wall exterior Page 15 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 The dose rates are developed based on the calculated maximum dose rates reported in UFSAR Table P.5-2 for the 24PTH-S-LC DSC and Table P.5-1 for the 24PTH-L DSC (the 24PTH-S DSC being bounded by the 24PTH-L DSC).

For the TC, a single set of dose rate limits is reported per DSC. As such, the dose rate limits correspond to the 24PTH-L DSC dose rates shown in UFSAR Table P.5-3 (and Table P.5-3a),

which bound those of the 24PTH-S-LC DSC shown in UFSAR Table P.5-5 (and Table P.5-5a).

The 24PTH TC dose rate limits are:

• 1200 mrem/hr, axial surface dose rate

• 1500 mrem/hr, radial surface dose rate The 24PTH-S-LC DSC loading is limited to HLZC #5 (maximum fuel assembly heat load of 1.5 kW), while the 24PTH-S and 24PTH-L DSC can be loaded with HLZC #2 (maximum fuel assembly heat load of 2.0 kW), per the CoC 1004 UFSAR Appendix B Amendment 18 Technical Specifications (TS).

Note that 10 CFR Part 72 does not specify dose rate limits for dry storage system surfaces nor at set distances from dry storage system surfaces. The Amendment 18 TS appropriately define the loading configurations for each subset DSC of the 24PTH DSC system and dose rate limits are appropriately developed in CoC 1004 Appendix A Amendment 18 ITE for the licensee to demonstrate compliance with 10 CFR 20, 10 CFR 72.104, and 10 CFR 72.106.

Impact:

No change as a result of this RAI.

Page 16 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.5:

Clarify the meaning of the phrase prior to installation as it applies to the minimum lead thickness specified for the DSC bottom shield plug on sheet 4 of 5 of Drawing No. NUH24PTH-1002-SAR, revision 3A.

On sheet 4 of 5 of Drawing No. NUH24PTH-1002-SAR, revision 3A, the drawing specifies a minimum lead thickness for the DSCs bottom shield plug. However, the specification includes the caveat that this is the minimum lead thickness prior to installation. The drawing should show the minimum lead thickness of the installed configuration. If upon installation the lead minimum thickness can be less than the currently specified value, the drawing should be revised to specify the installed lead minimum thickness. Also, the shielding analysis should be revised to use the installed lead minimum thickness for determining dose rates.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.5:

The lead thickness of the DSCs bottom shield plug accounted for in the analysis is based on the minimum lead thickness value defined in the SAR drawings. Therefore, Drawing NUH24PTH-1002-SAR has been revised to remove the note prior to installation in order to define the minimum lead thickness as installed.

Impact:

UFSAR Drawing NUH24PTH-1002-SAR, Sheet 4 of 5 has been revised as described in the response.

Page 17 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.6:

Confirm the shielding calculations for the DSC with the baskets uses an appropriate density for the high-strength low-alloy steel (HSLA) of the basket, revising the calculations as necessary.

The density of the steel components of the DSC and its basket can have a significant impact on the dose rates for the DSC in the different configurations in the transfer cask and storage module. It is because of density differences that bounding analyses for systems that include options to use different steels (stainless steel and carbon steel options) will evaluate bounding dose rates with the option having the lower density (approximately 7.85 grams per cubic centimeter (g/cm3) for carbon steel versus 7.92 g/cm3 for stainless steel). The HSLA steel used for the Type 3 basket can have an even lower density of about 7.75 g/cm3. Thus, with the basket having an impact on radial and axial dose rates, the shielding analysis models should use this density to determine dose rates for the 24PTH DSCs with this new basket. It is not clear from the application that the analysis models use an appropriate density for the HSLA steel of the basket. Sample input files indicate that the density associated with stainless steel was used.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.6:

The shielding analysis of a dry storage system involves the development of models that are expected to be consistent with the design, including dimensions and materials. When possible and justifiable, assumptions that simplify the model while preserving the overall conservatism of the evaluation may be employed. For example, it may be acceptable to simplify the model by smearing fuel assembly source and basket material in the packaging cavity. TN America LLCs (TNs) approach is to model the basket geometry while smearing fuel assembly material within the compartments. Under this approach, detailed elements of the basket may be simplified; for example borated neutron absorber material sheet is modeled as aluminum, steel plate material is modeled as standard stainless steel, etc.

High-strength low-alloy steel is used in the 24PTH Type 3 basket for its better mechanical properties and greater resistance to corrosion. The density of HSLA steel shown in UFSAR Table P.3.3-10 is 0.283 lb/in3 (or 7.83 g/cm3), which is approximately 1% lower than stainless steel density of 7.92 g/cm3. The effect of this 1% difference in density in the basket plates is not expected to be notable on dose rates that are more sensitive to the bounding source specification and the neutron and gamma shielding materials outside the basket. UFSAR Section P.5.4.6.2 and Section P.5.4.6.4 have been revised to clarify the basket steel plate material in the model.

Impact:

UFSAR Sections P.5.4.6.2 and P.5.4.6.4 have been revised as described in the response.

Page 18 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.7:

Confirm that the only gaps in the DSC shield plugs with the new lead option are radial gaps (i.e.,

that there are no axial gaps too). If axial gaps are allowed, revise the drawings to show the maximum axial gaps.

The revised drawings for the DSC (Drawing No. NUH24PTH-1001-SAR, rev 7A and Drawing No. NUH24PTH-1002-SAR, rev 3A) only show radial gaps for the new lead option in the 24PTH-S-LC DSCs shield plugs (top and bottom). However, section P.4.12.3 discusses axial gaps between the lead and steel shield plug components. The drawings should show the maximum dimensions of any axial gaps. Both axial and radial gaps can be important for shielding, particularly in understanding how much the lead may deform, including under drop accident conditions, and the impacts for dose rates that deformation will have.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.7:

A lead disc insert option is provided as an alternate to the lead pouring option for the 24PTH-S-LC DSC with a Type 3 basket. The axial gap mentioned in Section P.4.12.3 is to account for potential contact gaps, i.e., surface roughness between lead discs and adjacent surfaces. This is specific to the thermal analysis and does not reflect any pre-defined/actual axial gap in the top and bottom shield plug assemblies.

Impact:

No change as a result of this RAI.

Page 19 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.8:

Modify CoC appendix B tables that show allowed combinations of burnup, enrichment, and cooling time to show that the enrichment is a minimum assembly average initial enrichment, not a maximum.

Appendix B contains several tables for the different DSCs (e.g., tables 1-3k through 1-3m, 1-3o, and 1-3p for the 24PTH DSC) that specify allowed fuel assembly burnup, enrichment, and cooling time combinations. These combinations are specified for decay heats and to limit the radiation source terms of the spent fuel contents. For radiation source terms, the enrichment needs to be specified as a minimum assembly average initial enrichment since the radiation source term increases with decreasing enrichment. However, several tables, including those for the 24PTH DSC, incorrectly state the enrichment as maximum enrichments. The tables should also be clear that the burnup is the maximum assembly average burnup.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.8:

Labels for enrichments in the fuel qualification tables (FQTs) in CoC 1004 Appendix B have been updated to Assembly Average Initial Enrichment.

Burn-Up (BU) in FQTs in CoC 1004 Appendix B are defined as Assembly Average Burnup in various notes associated with FQTs.

Note that during fuel qualification, licensees are responsible for applying adequate uncertainties to fuel assembly enrichment (minimizing assembly average initial enrichment value) and burnup (maximizing assembly average burnup value).

Impact:

CoC 1004 Appendix B Tables 1-3i, 1-3k, 1-3m, 1-3n, 1-3o, 1-7k, and 1-7m have been revised as described in the response.

Page 20 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.9:

Provide an evaluation of lead slump or creep in the DSC top and bottom shield plugs and the shielding impacts of lead slump or creep for the following cases:

a. slump due to accident conditions (e.g., a side drop of the DSC in the transfer cask from the maximum height the transfer cask would be lifted in this configuration)

b. creep of the lead over the operational life of the DSC With radial gaps (and possibly axial gaps; see RAI 5.8), between the lead and the steel components of the shield plugs, there is potential for the lead shielding to deform either by slumping under accident conditions or with creep over time. This deformation of the lead shielding, depending on the extent possible, can have impacts on dose rates that are significant.

The application does not appear to include an analysis addressing slump under accident conditions or creep over time as the DSC rests in the storage module. For creep over time, while the radiation source term also decays over time, it should be shown that the source term decay is at least enough to offset the effects on shielding capability of the lead shield plugs due to lead creep.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.9:

Item a:

As indicated in UFSAR Sections P.5.4.6.4 and P.5.4.8, the accident dose evaluation assumes the cask neutron shield (either water or NS-3) and the neutron jacket (outer steel) to be lost.

The effect on neutron and gamma dose rates resulting from these assumptions is much more severe than a potential localized lead slump after a drop as this would affect mainly top or bottom gamma dose rates. Note that lead is not expected to vanish after a drop accident. The accident dose received by a person located at 100 meters, nearest boundary of the controlled area per 10 CFR 72.106, from the NUHOMS 24PTH system installation is not expected to increase substantially due to lead slump. The dose received by a person located at 100 meters, evaluated in UFSAR Section P.11.2.5.3 based on the accident dose analysis in UFSAR Section P.5.4.8, is adequate and the exposures are well within the limits of 10 CFR Part 72 for an accident condition.

Page 21 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Item b:

Lead creep is characterized by a potential continuing deformation of the metal under a steady load and/or increase in temperature. During storage, temperature decrease and source decay over time are expected to offset any potential lead creep. The design basis source for the HSM-H evaluation is determined to be 41 GWd/MTU, 3.3 wt% U-235 and 3 years cooling time corresponding to a decay heat of 2.0 kW per fuel assembly. CoC 1004 Amendment 18 TS Table 1-3k shows the decay heat of this burnup/enrichment combination is 1.5 kW per fuel assembly (492 kgU) after 3.7 years cooling time, i.e., a 25% reduction in decay heat in just 0.7 additional year cooling time. UFSAR Table P.5-24 and Table P.5-26 provide the expected decrease in dose rates for this burnup/enrichment combination from 2.0 kW per fuel assembly (3 years cooling time) to 1.5 kW per fuel assembly (3.7 years cooling time). The HSM-H dose rate at the roof drops from 7.9 mrem/hr to 5.1 mrem/hr, i.e., approximately 35% reduction in dose rate in just 0.7 additional year cooling time. In short, in less than 1 year, the decay heat and HSM dose rate are expected to drop, respectively, by 25% and 35%. The reduction in temperature and radiological source are expected to outpace the effect due to lead creep, if any. Furthermore, note that the HSM-H roof and front vent dose rates are the largest contributors to the site dose as the front door (and back) HSM-H dose rate is very small in comparison.

Impact:

No change as a result of this RAI.

Page 22 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.10:

Provide clarification that the geometric configuration of damaged fuel does not change under accident conditions or explain how a change in the configuration is bounded by the current shielding analysis.

The shielding analysis appears to treat the damaged fuel as intact fuel. The geometric configuration of the spent fuel contents impacts the geometric distribution of the source term that can lead to increased dose rates. Page P.5-8 of the updated final safety analysis report (UFSAR) indicates that damaged fuel has no impact on dose rates. The basis and justification for that statement are not clear.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.10:

Damaged fuel is defined in CoC 1004 Appendix B Amendment 18 Technical Specifications (TS)

Table 1-1l as assemblies containing missing or partial fuel rods, fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly, including non-cladding damage, is to be limited such that the fuel assembly is able to be handled by normal means. Missing fuel rods are allowed. The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions. As such, under normal and off-normal conditions, damaged fuel has no impact on dose rates as the source term would not be affected and gross axial source redistribution is not likely.

The potential impact of damaged fuel under accident conditions has been addressed, for example, in UFSAR Chapter U.5 for the 32PTH1 system (and Chapter T.5 for the 61BTH system) including assuming the damaged fuel turns to rubble. The accident scenario assumes that the neutron shield and steel neutron shield jacket (outer skin) have been torn off (see Section U.5.4.8 and Section U.11.2.5.3). While local peak dose rates are observed, the dose rates at a far distance (dose rates employed for the accident dose evaluations) are bounded by those calculated assuming intact fuel in damaged fuel locations. A similar conclusion is expected for the 24PTH system; therefore, the accident dose calculations documented in UFSAR Section P.11.2.5.3 are appropriately evaluated for all fuel conditions and show significant margin to the 10 CFR Part 72 limits for accident conditions.

UFSAR Section P.5.4.8 has been updated to reference the damaged fuel evaluation in the accident condition performed in Chapter U.5.

Impact:

UFSAR Section P.5.4.8 has been revised as described in the response.

Page 23 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.11:

Revise the tables and figures in chapter P.5 of the UFSAR that describe the analysis model specifications to include the changes to the analysis models for this amendment.

While the text of chapter P.5 discusses the differences (e.g., lead density), the tables and figures in chapter P.5 are an important source for understanding the models quickly and easily.

Thus, they should also reflect the information for analysis models with the new basket (e.g.,

density for the HSLA, configuration of basket materials) and the lead in the 24PTH-S-LC DSCs top and bottom shield plugs (e.g., lead density, lead dimensions). Consistent description of the models for the different models is necessary to ensure correct and consistent understanding of the analyses.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.11:

A note has been added to UFSAR Table P.5-14 to indicate that the lead density in top and bottom shield plug models of the 24PTH-S-LC Type 3 is 11.00 g/cm3. The lead thicknesses are shown in Drawings NUH24PTH-1002-SAR and NUH24PTH-1001-SAR, respectively, (i.e., 4.30 inches for the top shield plug and 2.50 inches for the bottom shield plug).

The other materials employed in the Monte Carlo N-Particle (MCNP) models are standard materials including high-strength low-alloy steel (HSLA) in the basket modeled as stainless steel; see the response to RAI 6.6.

Impact:

UFSAR Table P.5-14 has been revised as described in the response.

Page 24 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.12:

Justify why the surface average dose rates on the HSM front do not increase with the changes in the basket for the 24PTH-L in the HSM-H and for the basket and shield plug changes for the 24PTH-S-LC in the HSM Model 102.

The maximum dose rates for the front surfaces change significantly (46% increase for the 24PTH-L and 45% increase for the 24PTH-S-LC in their respective HSM). Given the changes to the basket and, for the 24PTH-S-LC, the shield plugs, and that the changes affect a significant surface area, the staff expects that the average dose rates will also increase. Thus, further information is needed to understand why the average dose rates are unchanged. Otherwise, the analyses for those dose rates should be modified to reflect the increases in dose rates due to the changes to the basket and shield plugs for the respective variants of the 24PTH DSC.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.12:

Updated Final Safety Analysis Report (UFSAR) Table P.5-1 Note 7 indicates that the HSM-H door exterior surface (centerline) dose rate for the 24PTH-L Type 3 is 1.9 mrem/hr, which is a 46% increase from 1.3 mrem/hr of the 24PTH-L Type 2. The contribution of the HSM-H door dose rate to the HSM-H front surface average dose rate is very limited since the latter is dominated by the dose rate exiting the front vents, since the 24PTH-L Type 2 HSM-H front birdscreen dose rate is 1237 mrem/hr. Furthermore, it should be noted that UFSAR Table P.5-3ashows a significant reduction in the cask side surface (radial) dose rate for the 24PTH-L Type 3 compared to that of the 24PTH-L Type 2 in Table P.5-3 (1.25 E+3 mrem/hr in Table P.5-3a Vs 1.50 E+3 mrem/hr in Table P.5-3). This is due to thicker transition rails in the 24PTH-L Type 3; therefore, the 24PTH-L Type 3 HSM-H front birdscreen dose rate is expected to be bounded by that of the 24PTH-L Type 2.

A similar observation can be made for 24PTH-S-LC in the HSM Model 102 using UFSAR Table P.5-2 Note 5, and the cask side surface (radial) dose rates in UFSAR Table P.5-4 and UFSAR Table P.5-4a.

Impact:

No change as a result of this RAI.

Page 25 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.13:

Confirm and justify the dose rate differences seen in the application for the transfer cask and HSM.

The radiation source terms and the geometry of those source terms for the spent fuel contents of the 24PTH-L and 24PTH-S-LC are not changing in this amendment. The amendment only changes the basket from the Type 2 to the Type 3 basket for both these DSC variants. The amendment also changes the lead shield plugs in the 24PTH-S-LC, reducing the lead density and reducing the minimum lead thickness in the bottom shield plug. Since only the basket is changing for the 24PTH-L, the staff expects that the impact to dose rates should be uniform for different analyzed configurations at the top and bottom axial ends of the DSC in the transfer cask and the storage module. The applicants results indicate this is the case with the exception of the transfer cask top dose rates. All transfer cask dose rates increase by 25 to 33% except the top transfer operations configuration, which only increases 4.6%. The staffs simple calculations indicate that the change across all configurations should be a uniform increase of over 20%. The staff notes the increase for the HSM-H door is higher than expected based on its simple calculations, but the staff can find that to be acceptable (as the applicants analyzed increase bounds the increase determined in the staffs evaluation).

For the 24PTH-S-LC, both the basket and the lead shield plugs are changing, with greater changes to the bottom lead shield plug (vs. the top lead shield plug). Yet, the applicants dose rate results indicate that the transfer operations configuration dose rates for the top decrease slightly. This is inconsistent with the source term and source term configuration not changing, but the shielding due to the basket and the lead shield plug decreasing. Also, a comparison of the differences in HSM door dose rates for the 24PTH-L and 24PTH-S-LC show the difference in dose rates is comparable between the two DSC variants (46% increase vs. 45% increase).

This is not consistent with differences in shielding changes between the two DSC variants. The relative increase in dose rates for the 24PTH-S-LC should be significantly higher (versus the increase seen for the 24PTH-L) given reduced shielding in not only the basket, but also in the lead shield plug (lower density and thinner minimum thickness). The staffs simple calculations indicate that the transfer cask top dose rates should increase significantly (by over 50%) and the transfer cask bottom dose rates and HSM door dose rates should also increase significantly (by 60 to 70%).

Based on the above, further justification and explanation of the dose rate changes is needed, or the analyses need to be corrected to appropriately reflect the shielding impacts due to the basket and lead shield plug changes. This is important for understanding and correctly characterizing dose rates for occupational dose assessments as well as normal, off-normal, and accident conditions doses for public dose assessments.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d) and ensure the system is sufficient to facilitate licensee compliance with 10 CFR 20.1101 and 20.1201(a) and 10 CFR 72.126(a)(6).

Page 26 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Response to RAI 6.13:

Updated Final Safety Analysis Report (UFSAR) Section P.5.4.7.3 provides the analysis of the 24PTH-L DSC in the OS197FC transfer cask (TC). Results for the 24PTH-L Type 2 and 24PTH-L Type 3 are shown, respectively, in UFSAR Table P.5-3 and Table P.5-3a. Comparison of the dose rates in UFSAR Table P.5-3 and Table P.5-3ashows that side dose rate decreases by approximately 17% (1.50E+3 mrem/hr for 24PTH-L Type 2 Vs 1.25E+3 mrem/hr for 24PTH-L Type 3), while the top axial dose rate increases slightly by 4.6% (2.61E+2 mrem/hr for 24PTH-L Type 2 Vs 2.73E+2 mrem/hr for 24PTH-L Type 3) and bottom axial dose rate increases by 31%

(4.23E+3 mrem/hr for 24PTH-L Type 2 Vs 5.55E+3 mrem/hr for 24PTH-L Type 3). The last paragraph of UFSAR Section P.5.4.7.3 provides an explanation of the differences as side dose rates decrease for the Type 3 basket compared to the Type 2 basket due to the different transition rail design. However, because the basket members are generally lighter in the Type 3 basket, the end dose rates increase for the Type 3 basket compared to the Type 2 basket.

Note that, as demonstrated in UFSAR Table P.5-9 through Table P.5-11, the effect of the basket change is less pronounced for the top dose rate since the top region source is lower than that of the bottom region; furthermore, the top nozzle ends well below the top shield plug as opposed to the bottom nozzle (the bottom nozzle sits just above the bottom shield plug), as shown in UFSAR Figure P.5-8.

UFSAR Section P.5.4.7.4 provides the analysis of the 24PTH-S-LC DSC in the Standardized TC. Results for the 24PTH-S-LC Type 2 and 24PTH-S-LC Type 3 are shown, respectively, in UFSAR Table P.5-5 and Table P.5-5a. Comparison of the dose rates in UFSAR Table P.5-5 and Table P.5-5ashows that side dose rate decreases by approximately 29% (5.77E+2 mrem/hr for 24PTH-S-LC Type 2 Vs 4.10E+2 mrem/hr for 24PTH-S-LC Type 3) when the top axial dose rate is relatively similar (3.25E+1 mrem/hr for 24PTH-S-LC Type 2 Vs 3.23E+1 mrem/hr for 24PTH-S-LC Type 3), and the bottom axial dose rate increases by 41% (4.75E+3 mrem/hr for 24PTH-S-LC Type 2 Vs 6.70E+3 mrem/hr for 24PTH-S-LC Type 3). The last paragraph of UFSAR Section P.5.4.7.4 provides an explanation of the differences as side dose rates decrease for the Type 3 basket compared to the Type 2 basket due to the different transition rail design. However, because the basket members are generally lighter in the Type 3 basket, the end dose rates increase for the Type 3 basket compared to the Type 2 basket. In addition, end dose rates also increase due to modeling the machined-lead shield plug option.

Note that the lead thickness in the top shield plug is the same between Type 2 and Type 3 baskets, while lead density employed in the Type 3 is reduced (11.34 g/cm3 in the Type 2 calculation and 11.00 g/cm3 in the Type 3 calculation). The overall effect on the top total dose rate reflects the minimum change, the difference in dose rate is within the MCNP uncertainty

(). The overall effect of the change on the bottom total dose rate is more significant as the lead thickness in the Type 3 is 2.50 inches, while it is 2.75 inches in the Type 2 basket.

Impact:

No change as a result of this RAI.

Page 27 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 6.14:

Explain how the additional allowed weight for the fuel assemblies and control components in the Type 3 basket will or will not affect the radiation source terms, including total source terms and axial distribution of source terms and modify the shielding analysis as necessary.

Since the Type 3 basket weighs less than the currently approved basket types in the 24PTH DSC, the applicant has proposed increasing the allowed weight of the fuel assembly and control component contents in each basket cell. The increase is about 15 kilograms. With an increase in mass comes an increase in source term as well. It is not clear how this increase in source term from increased allowed mass has been considered in the shielding analyses. The maximum mass of uranium per fuel assembly remains fixed. So, the increased mass would be in control components or assembly hardware, which increases the cobalt-60 source due to activation of these items. While increased mass does increase self-shielding, the staffs simple calculations indicate that the increased self-shielding is not sufficient to compensate for the increase in source term (i.e., dose rates will increase). Additionally, the distribution of the added weight and added source term can have a significant impact on dose rates. If the added mass is in either the upper or lower hardware zones of the assemblies, the respective axial dose rates for the DSC will increase, potentially significantly.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 6.14:

Change no. 1 in Enclosure 2 of submittal E-60447, which relates to the increase of the NUHOMS-24PTH Dry Shielded Canister (DSC) weight limit from 1682 lb to 1715 lb for fuel assembly plus control component (CC) based on the lighter Type 3 basket, does not allow loading of higher overall source in a Type 3 DSC. The maximum initial uranium content is still limited to 492 kg per assembly as indicated in CoC 1004 Appendix B Amendment 18 Technical Specifications (TS) Table 1-1I, and the maximum CC source per DSC is limited to values shown in CoC 1004 Appendix B Amendment 18 TS Table 1-1n.

For loading fuel assemblies in a NUHOMS-24PTH Type 3 DSC, the licensee is expected to demonstrate, among other requirements in CoC 1004 Appendix B Amendment 18 TS Table 1-1I, that the maximum initial uranium content is less than 492 kg per assembly, the fuel assembly weight with CC is less than 1715 lb, and the total CC source per DSC is lower than the values shown in CoC 1004 Appendix B Amendment 18 TS Table 1-1n.

Impact:

No change as a result of this RAI.

Page 28 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Radiation Protection:

RAI 10.1:

Provide further clarification and information to explain the basis for the occupational dose estimate being only slightly impacted by the increased dose rates on the DSC top, transfer cask, and storage module that result from the DSC and basket changes.

In section P.10.1 of the application, the applicant states that the net effect of the changes on the occupational exposure is an increase of only 3%. Given the DSC and basket changes the staff estimates, using the applicants reported dose rate information, that the impact could be more substantial. However, to confirm the conclusion, the staff needs information regarding personnel positions for the different operations covered by the table P.10-1 method for estimating occupational exposures, including where those positions are versus the dose rates calculated in chapter P.5 for the different configurations of the DSC in the transfer cask and the storage module.

The staff recognizes that the applicant also provided some limited data and a short discussion regarding measured occupational exposures from actual loading campaigns. However, the data and discussion do not include any meaningful information for applying the provided data to support the applicants determination that the unchanged table P.10-1 exposure estimates remain adequate and conservative for the modified DSC and basket.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d) and ensure the system is sufficient to facilitate licensee compliance with 10 CFR 20.1101 and 20.1201(a) and 10 CFR 72.126(a)(6).

Response to RAI 10.1:

The TN Americas LLC (TN) methodology for evaluating the occupational dose has been consistent for all storage systems licensed under CoC 1004 throughout the amendments. The methodology combines average dose rates in the vicinity of the operations, and assumed personnel and duration involved in the operations. Note that average dose rates are appropriately employed for the occupational dose evaluation to reflect the personnel movements in the vicinity of operations. As low as reasonably achievable (ALARA) procedures would prevent personnel from standing still at a high dose rate location for an extended period.

This approach for the updated final safety analysis report (UFSAR) occupational dose exposure evaluation has shown significant conservatism over the amendments compared to actual loading occupational dose exposure.

Table RAI 10.1-1 below provides the average dose rates for Type 2 and Type 3 dry shielded canisters (DSCs) used in the occupational dose exposure evaluation. As mentioned in UFSAR Section P.10.1, the 24PTH-L Type 3 DSC has lower cask side dose rates and higher cask end dose rates than the 24PTH-L Type 2 DSC. Table RAI 10.1-2 provides the comprehensive occupation dose exposure evaluation for the Type 3 DSC in the similar format to UFSAR Table P.10-1 for the Type 2 DSC. The total dose exposure for Type 3 is 4609 person-mrem, which corresponds to an increase of 3% compared to that of Type 2 (4475 person-mrem, UFSAR Table P.10-1) as indicated in UFSAR Section P.10.1. For conciseness, the complete occupation dose exposure evaluation for the Type 3 DSC is not reported in the UFSAR, which only indicates the 3% increase.

Page 29 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Table RAI 10.1-1: Dose Rates for Dose Exposure Evaluation 24PTH-L Type 24PTH-L Type 2 (mrem/hr) 3 (mrem/hr)

Ambient 2 2 Decontamination Side 1 foot average 371 292 Side 3 feet average 238 186 Top 1 foot average < 90 cm 466 566 Welding Side 1 foot average 382 317 Side 3 feet average 243 200 Top 1 foot average < 90 cm 557 646 Top 3 feet average < 90 cm 362 416 Transfer Side 1 foot average 475 397 Side 3 feet average 304 252 Top 1 foot average < 90 cm 59 68 Top 3 feet average < 90 cm 36 40 Bottom 3 foot average < 90 cm 361 446 HSM Front 10.4 10.4 Page 30 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Table RAI 10.1-2: Occupation Dose Exposure Evaluation, Type 3 Basket Area Total Dose Exposure

1. of Duration Rate (person-Location Task Description workers (hr) (mrem/hr) mrem)

Place the DSC into the Transfer 3 1 2 6 Cask Fill the Cask/DSC Annulus with Auxiliary Building and Fuel Pool Clean Water and Install the 2 2 2 8 Inflatable Seal Fill the DSC Cavity with Water 1 6 2 12 (borated for PWRs)

Place the Cask Containing the DSC 5 0.5 2 5 in the Fuel Pool Verify and Load the Candidate Fuel 3 5 2 30 Assemblies into the DSC Place the Top Shield Plug on the 2 1 2 4 DSC 5 0.5 2 5 Remove the Cask/DSC from the Fuel Pool and Place them in the 1 0.033 292 10 Decon Area 1 1 186 186 Decontaminate the Outer Surface of 1 1.75 186 326 the Cask 1 1 2 2 Decontaminate the Top Region of 2 0.5 566 566 the Cask and DSC 1 1 2 2 Drain Water from the DSC 1 0.083 292 24 1 0.25 646 161 Remove Cask/DSC Annulus Seal Cask Decontamination Area 1 1.25 416 520 and Set-Up Welding Machine 1 1.5 2 3 Weld the Inner Top Cover to the 2 6 2 24 DSC Shell and Perform NDE (PT) 1 0.5 416 208 1 0.25 416 104 Drain the Cask/DSC Annulus and 1 0.017 646 11 the DSC Cavity 1 0.5 2 1 Vacuum Dry and Backfill the DSC As above 116 with Helium Helium Leak Test the Shield Plug 2 1 2 4 Weld Seal Weld the Prefabricated Plugs to the Vent and Siphon Port and 1 0.5 416 208 Perform NDE (PT)

Page 31 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Table RAI 10.1-2: Occupation Dose Exposure Evaluation, Type 3 Basket (continued)

Total Area Dose Exposure

1. of Duration Rate (person-Location Task Description workers (hr) (mrem/hr) mrem) 1 0.5 2 1 Fit-Up the DSC Top Cover Plate 1 0.5 416 208 1 1.25 416 520 Weld the Outer Top Cover Plate to 1 1.5 2 3 DSC Shell and Perform NDE (PT) 2 14 2 56 1 0.5 416 208 Install The Cask Lid 2 1 40 81 Ready the Cask Support Skid and Reactor/Fuel 2 2 2 8 Transport Trailer for the Service Building Bay Place the Cask onto the Skid and 2 0.5 252 252 Trailer Secure the Cask to the Skid 1 0.25 252 63 Ready The Cask Support Skid and 2 2 negligible 0 Transport Trailer for the Service Transport the Cask to ISFSI 6 1 negligible 0 Position the Cask in Close 3 1 negligible 0 Proximity with the HSM Remove the Cask Lid 2 1 51 102 ISFSI Site Align and Dock the Cask with the 2 0.25 263 131 HSM Position and Align Ram with Cask 2 0.5 263 263 Remove Ram Access Cover Plate 1 0.25 446 111 Transfer the DSC from the Cask to 3 0.5 negligible 0 the HSM Lift the Ram Back onto the Trailer 2 0.083 263 44 and Un-Dock the Cask from the Install HSM Access Door 2 0.5 10.4 10 Totals N/A 63 N/A 4609 Page 32 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 Impact:

No change as a result of this RAI.

Page 33 of 39

RAIs and Responses - Public Enclosure 3 to E-61864 RAI 10.2:

Provide an evaluation of the impact of the changes to the DSC and basket on the evaluation in the UFSAR on the demonstration for meeting the annual dose limits in 10 CFR 72.104(a), either modifying the evaluation for 10 CFR 72.104(a) or justifying that the current evaluation bounds the system with the modified DSC and basket.

With dose rates increasing as a result of the DSC and basket changes, the staff expects that there will be impacts to the evaluations to support demonstration of meeting the 10 CFR 72.104(a) limits. However, the applicant has not provided revised evaluations in chapter P.10 of the UFSAR to address this. Nor has the applicant provided justification for why the current analyses are bounding for the system with the modified DSC and basket.

This information is needed to confirm compliance with 10 CFR 72.236(b) and (d).

Response to RAI 10.2:

The off-site site dose evaluation is presented in Updated Final Safety Analysis Report (UFSAR)

Section P.10.2 for two generic independent spent fuel storage installation (ISFSI) layouts containing design-basis fuel. The first evaluated generic ISFSI is a 2x10 back-to-back array of HSM-Hs loaded with design basis fuel in NUHOMS 24PTH-L Dry Shielded Canisters (DSCs).

The second evaluated generic layout is a two 1x10 front-to-front arrays of HSM-Hs loaded with design-basis fuel in NUHOMS 24PTH-L DSCs. A similar evaluation is also performed for the NUHOMS 24PTH-S-LC DSCs in HSM-Model 102s.

As described in UFSAR Section P.10.2, the base model for the array of HSM-Hs/HSM-Model 102s is a box enveloping the array. As such, average surface dose rates (roof, sides, back and front) determined in UFSAR Table P.5-1 for NUHOMS 24PTH-L DSC in HSM-H and UFSAR Table P.5-2 for NUHOMS 24PTH-S-LC DSCs in HSM-Model 102 are employed as inputs for the side dose evaluations. As these average surface dose rates are shown to be bounding for all DSC Types, i.e., Type 2 average surface dose rates bound Type 3 average surface dose rates for both NUHOMS 24PTH-L DSC in HSM-H and NUHOMS 24PTH-S-LC DSC in HSM-Model 102, the generic ISFSI evaluations remain unchanged.

Impact:

No change as a result of this RAI.

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RAIs and Responses - Public Enclosure 3 to E-61864 Operation Procedures and Systems:

RAI 11.1:

Describe the mechanism to alert general licensees of the need to review all design basis parameters that may be affected by adverse weather conditions (e.g., operating windspeed for any handling equipment used) and consider these design basis parameters along with weather forecasts for decision making and implementation when establishing these weather-related administrative controls for the expected duration of the activity.

While an acceptance criterion of "tornado watches, advisories, or warnings may protect Structures Systems and Components (SSCs) from design basis tornado accidents, it may not alone protect SSCs from all normal and off normal wind conditions and therefore additional evaluation may be necessary to demonstrate an analyzed wind speed that is appropriate for normal operation of SSCs. The purpose of the weather-related administrative controls should be to prevent the DSC and any DSC ancillary systems from exceeding their design basis parameters due to weather conditions during handling operations. Design basis parameters may be exceeded by weather conditions that do not result in a weather alert.

This information is needed to confirm compliance with 10 CFR 72.236(l).

Response to RAI 11.1:

This Amendment 18 scope item is meant to address amendments to be requested, as discussed in Action 3. b) of Enforcement Guidance Memorandum 22-001, Enforcement Discretion for Noncompliance of Tornado Hazards Protection Requirements at Independent Spent Fuel Storage Installations (EGM 22-001).

EGM 22-001 addresses administrative controls for severe hazards weather advisories, watches, or warnings related to tornado hazards. Therefore, the administrative controls included in the UFSAR changes for Amendment 18 are limited to tornado watches, advisories, or warnings.

If SSCs have other weather-related limitations, such as wind speed, those must be addressed in the related design and licensing basis, and the related design and licensing basis documents should so stipulate.

Impact:

No change as a result of this RAI.

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RAIs and Responses - Public Enclosure 3 to E-61864 RAI 11.2:

Provide the basis for checking the forecast for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Specifically provide the basis that 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> bounds all expected handling operations. Alternatively, specify that the weather should be checked for the expected duration of the handling operation.

The licensee should determine the expected bounding duration of the handling operation by either benchmarking or dry runs. The duration should then be periodically assessed based upon operating experience.

This information is needed to confirm compliance with 10 CFR 72.236(l).

Response to RAI 11.2:

UFSAR Section 5.1.1.5 has been revised to indicate that the licensee should determine the expected bounding duration of the handling operation by either benchmarking or dry runs, including a margin for contingency. The weather would then be checked for the expected duration of the handling operation and, subsequently, that duration would then be periodically assessed based on operating experience.

Impact:

UFSAR Section 5.1.1.5 has been revised as described in the response.

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RAIs and Responses - Public Enclosure 3 to E-61864 RAI 11.3:

Provide additional guidance on how licensees will obtain weather forecasts prior to each loading/unloading evolution.

Licensees may use information from the National Weather Service unless another resource for the site (e.g., the National Oceanic Atmospheric Administration, Weather Forecast Office nearest the site, weather channel, etc.) is already approved for use or can be justified as providing equivalent information in terms of timeliness and accuracy. Prescribe specific times that the weather forecast would be checked, the area(s) considered, and the frequency of the forecast checks accounting for the site configuration and time necessary to be able to bring the system into an analyzed configuration at any time, if necessary, when there is a change in forecast. Describe other conditions that cannot be met (e.g., occurrence of significant delays, equipment malfunctions, or manpower/labor issues) that would necessitate having to place the storage system SSCs in a safe and analyzed condition.

This information is needed to confirm compliance with 10 CFR 72.236(l).

Response to RAI 11.3:

TN agrees that licensees should use the National Weather Services hazardous weather outlook information unless another resource for the site is already used and can be justified as providing equivalent information in terms of timeliness and accuracy. UFSAR Section 5.1.1.5 has been revised to add this guidance.

Although UFSAR Section 5.1.1.5 currently has some of the aspects regarding times to check the forecast, areas considered, etc., mentioned in the RAI, Section 5.1.1.5 has been revised to include all of these items.

Regarding other conditions that cannot be met, although malfunctions and unexpected delays during short-duration outdoor activities can and do occur, they are infrequent and the likelihood of a malfunction or delay occurring at the same time as the unexpected loss of an acceptable weather forecast is felt to be so low that it should be considered non-credible. Such malfunctions and delays should be handled on a case-specific basis and within the licensees corrective action program.

Impact:

UFSAR Section 5.1.1.5 has been revised as described in the response.

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RAIs and Responses - Public Enclosure 3 to E-61864 RAI 11.4:

Provide a description of meteorological monitoring criterion to be satisfied prior to the start of the ISFSI transient operations.

Identify the actions to be performed before commencing ISFSI transient operations (e.g.,

notification of control room, assigning personnel to monitor weather during ISFSI transient operations, site walkdowns that identify and secure any potential hazards, etc.) that can preclude short term operations during periods of adverse weather or during periods when adverse weather is predicted. Satisfactory completion of these criteria should be recorded and maintained with the documentation for the DSC campaign and should be checked no greater than the frequency established in Step 1 thereafter until ISFSI handling operations are completed.

This information is needed to confirm compliance with 10 CFR 72.236(l).

Response to RAI 11.4:

The aspects discussed in this RAI have been incorporated into UFSAR Section 5.1.1.5.

Impact:

UFSAR Section 5.1.1.5 is revised as described in the response.

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RAIs and Responses - Public Enclosure 3 to E-61864 RAI 11.5:

Provide a basis for not applying administrative controls to infrequently performed maintenance and inspection related activities during which ISFSI important to safety SSCs are in an unanalyzed condition.

Since performance of aging management activities may also require placing SSCs into unanalyzed conditions, consideration should also be given to applying administrative controls to these activities as well.

This information is needed to confirm compliance with 10 CFR 72.236(l).

Response to RAI 11.5:

Any planned activities, including maintenance, aging management inspections, etc., where the configuration of important-to-safety structures, systems, and components (SSCs) is not analyzed for tornado hazards, are subject to these administrative controls.

UFSAR Section 5.1.1.5, which provides the detailed administrative controls, has been revised to make this clear. UFSAR Section 3.2.1, Tornado and Wind Loadings, has also been revised to make this clear.

Impact:

UFSAR Sections 3.2.1 and 5.1.1.5 have been revised as described in the response.

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