ML22216A078

Request for Exemption from Various 10 CFR Part 72 Regulations Resulting from Non-Destruction-Examination Compliance

ML22216A078
Person / Time
Site:	Sequoyah, 07201032
Issue date:	08/04/2022
From:	Marshall T Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML22216A077	List: ML22216A078
References
Download: ML22216A078 (17)

Text

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 August 4, 2022 10 CFR 72.7 ATTN: Document Control Desk Director, Division of Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034

Subject:

Sequoyah Nuclear Plant Request for Exemption from Various 10 CFR Part 72 Regulations Resulting from Non-Destruction-Examination Compliance

Reference:

TVA letter to NRC dated February 24, 2022, Sequoyah Nuclear Plant Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2)

(Agencywide Documents Access and Management System [ADAMS] Accession No. ML22059B061)

Pursuant to 10 CFR 72.7, Specific exemption, TVA requests an exemption from the requirements of 10 CFR 72.212(a)(2), 72.212(b)(3), 72.212(b)(5)(i), 72.212(b)(11) and 72.214.

The regulations require, in part, compliance to the terms, conditions, and specifications of the Certificate of Compliance (CoC) and amended CoC listed in §72.214. This exemption request from the aforementioned regulations is specific to conditions of the Holtec International, Inc. Certificate Number 1032 (CoC-1032). The condition of CoC-1032 Appendix B, Section 3.3, "Codes and Standards," specifies American Society of Mechanical Engineers Boiler and Pressure Vessel Code, 2007 Edition as the governing Code for the HI-STORM FW System. The Code requires that 100 percent of the weld seam joining shell-to-shell of the multiple-purpose canister (MPC) be inspected by radiography test.

During a review of manufacturing documents, Holtec identified that MPC-37 Serial Number 234 at Sequoyah Nuclear Plant (SQN) Independent Spent Fuel Storage Installation (ISFSI) has a MPC longitudinal shell-to-shell weld that may not have been properly digitally radiographed after a weld repair. The missing radiograph is for approximately 7.5 inches of weld repair located between 14 and 25 inches from the bottom of the MPC baseplate. MPC-37 Serial Number 234 was already in use and was registered on February 4, 2022, in the Reference letter.

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 U.S. Nuclear Regulatory Commission Page 2 August 4, 2022 provides the Exemption Request including the need and justification for the issuance of an exemption. Enclosure 2 to this letter is an affidavit prepared in accordance with 10 CFR 2.390 requesting that the proprietary evaluation provided in Enclosure 3 be withheld from public disclosure. provides the Certificate Holders (Holtec International, Inc.) proprietary evaluation of the condition relative to the HI-STORM FW System licensing basis that includes SQN site specific conditions for a loaded MPC.

When separated from Enclosure 3, this cover letter, Enclosures 1, and 2 are decontrolled.

This document contains no new regulatory commitments.

If you have any questions, please contact Jeff Sowa, Site Licensing Manager, at (423) 843-8129.

Respectfully, Thomas Marshall Site Vice President Sequoyah Nuclear Plant

Enclosures:

1. Specific Exemption Evaluation

2. Affidavit Pursuant to 10 CFR 2.390 to Withhold Information from Public Disclosure

3. Response to Request for Technical Information RRTI-3087-007R2 (Proprietary) cc:

NRC Regional Administrator - Region II NRR Project Manager - Sequoyah Nuclear Plant NRC Senior Resident Inspector Sequoyah Nuclear Plant Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

INDEPENDENT SPENT FUEL STORAGE INSTALLATION Specific Exemption Evaluation

1. Introduction and Background The Holtec International, Inc., (Holtec) HI-STORM FW dry cask storage (DCS) system is designed to hold and store spent fuel assemblies for independent spent fuel storage installation (ISFSI) deployment. The system is listed in 10 CFR 72.214, List of approved spent fuel storage casks, as Certificate Number (No.) 1032 herein referred to as CoC-1032. This DCS system is used by the Tennessee Valley Authority (TVA) at SQN, in accordance with 10 CFR 72.210, "General license issued." SQN uses the multipurpose canister (MPC) that holds up to 37 pressurized water reactor fuel assemblies.

Between January 7 and February 11, 2022, SQN placed into use six HI-STORM FW spent fuel casks at the SQN ISFSI in accordance with Amendment 3 of Holtecs CoC-1032.

Condition No. 10 of CoC-1032, in part, authorizes the DCS system for general license use subject to the conditions specified by 10 CFR 72.212, Conditions of general license issued under §72.210, the certificate, and the certificate Appendices A and B. Condition No. 6 of CoC-1032 states, "Features or characteristics for the site or systems must be in accordance with Appendix B to this certificate." CoC-1032 Appendix B, Approved Contents and Design Features, Section 3.3 "Codes and Standards, provides the information on American Society of Mechanical Engineers (ASME) Code requirements for the MPC.

During a root cause extent of condition investigation at Holtec, a review of manufacturing documents identified that MPC-37 Serial Number 234 has a longitudinal shell-to-shell weld for which no digitally radiographic examination (RT) is available following a weld repair.

The weld repair was approximately 16.5 inches in length beginning between 8.5 and 25 inches from the bottom of the MPC baseplate. An RT examination is not available for the weld repair length of approximately 7.5 inches between 14 and 25 inches from the bottom of the MPC baseplate.

2. Request for Exemption In accordance with 10 CFR 72.7, "Specific exemptions," TVA SQN is requesting an exemption from the following requirements, which will enable SQN to continue to use MPC Serial Number 234 in its current condition:

10 CFR 72.154(b), which states in part "The licensee, applicant for a license, certificate holder, and applicant for a CoC shall have available documentary evidence that material and equipment conform to the procurement specifications prior to installation or use of the material and equipment."

10 CFR 72.212(a)(2), which states "This general license is limited to storage of spent fuel casks approved under the provisions of this part."

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-1 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 10 CFR 72.212(b)(3), which states the general licensee must "Ensure that each cask used by the general licensee conforms to the terms, conditions, and specifications of a CoC or an amended CoC listed in§ 72.214."

10 CFR 72.212(b)(5)(i), which states "The cask, once loaded with spent fuel or once the changes authorized by an amended CoC have been applied, will conform to the terms, conditions, and specifications of a CoC or an amended CoC listed in §72.214".

The portion of 10 CFR 72.212(b)(11) which states that "The licensee shall comply with the terms, conditions, and specifications of the CoC . . . "

10 CFR 72.174 in part requires that Records pertaining to the design, fabrication, erection, testing, maintenance, and use of structures, systems, and components important to safety must be maintained by or under the control of the licensee or certificate holder until the NRC terminates the license or CoC.

10 CFR 72.214 lists CoC-1032 as an approved cask for storage of spent fuel under the condition specified in the CoC.

Regulations 10 CFR 72.212(a)(2), 72.212(b)(3), 72.212(b)(5)(i), 72.212(b)(11), and 72.214 involve conforming to, or complying with, the terms, conditions, and specifications of a CoC or an amended CoC. Regulations 10 CFR 72.154(b) and 72.174 pertain to maintaining records.

The exemption request is for MPC Serial Number 234 which does not fully comply with CoC-1032 requirements:

Condition No. 6 of CoC-1032 which states, "Features or characteristics for the site or systems must be in accordance with Appendix B to this certificate."

Appendix B, Section 3.3 of the CoC which states in part that the ASME Boiler and Pressure Vessel Code, 2007 Edition is the governing code for the HI-STORM FW MPCs, with certain approved alternatives.

Specifically, MPC Serial Number 234, does not comply with CoC-1032 Appendix B, Section 3.3 for ASME Section III NB-5000 because SQN does not have a document of an RT examination following a weld repair. ASME Section III, Division 1 Subsection NB are the rules for construction of nuclear facility Class 1 components.

3. Justification The Holtec HI-STORM FW DCS system (i.e., a loaded MPC, stored within a HI-STORM overpack) provides criticality control, shielding, heat removal, and confinement functions, independent of any other facility structures or components. The structural design of the cask system also maintains the integrity of the fuel during storage.

Summary of Safety Analysis Discipline Safety Analysis Structural Safety margins to the HI-STORM FW FSAR acceptance criteria are maintained for the updated analysis Confinement Confinement integrity is maintained, as the structural analysis demonstrates Criticality No impact to criticality analysis Shielding No impact to shielding analysis Thermal No impact to thermal analysis The MPC design and certification is based on compliance with ASME Boiler and Pressure Vessel Code Section III, with certain alternatives already reviewed and approved by the Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-2 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 NRC. Portions of ASME Section III (NB-5000) require that weld repairs be inspected to the same criteria as the initial welds. The MPC longitudinal shell-to-shell weld is inspected by digital RT in accordance with Section III, Subsection NB.

The weld length that cannot be confirmed by documentary evidence to be radiographed is approximately 7.5 inches out of the approximately 690 inches (1.1%) of the MPC welds that were confirmed to be fully inspected. It is important to note that the weld repair was performed in accordance with all quality procedures. The condition for which this exemption is requested is only related to the post repair volumetric examination.

Therefore, although the weld is expected to meet all strength requirements, the structural analysis assumes a weld strength reduction factor for conservatism. The following discussion demonstrates that the MPC continues to meet all its design basis requirements and safety functions.

Structural The longitudinal weld seam joining the shell in the subject MPC is assumed to have less than 100% provable joint efficiency because a repaired weld mass was not properly digitally radiographed after a weld repair. To account for the missing volumetric examination, the structural analysis was revised by applying a reduction in the weld joint efficiency of the longitudinal shell weld (classified as Category A weld joint in the ASME Code, Subsection NB) and its effect on the welds capacity is quantified to ensure that the safety margins remain positive.

HI-STORM FW FSAR [Reference 1] specifies the governing load combination for the MPC longitudinal weld pertaining to the short-term normal, off-normal and accident conditions.

These load combinations include combined action of the internal pressure and the differential thermal expansion resulting from the temperature contour in the shell for short-term normal and off-normal conditions.

ASME Section III, NB does not discuss weld joint efficiencies, hereafter referred to as stress-reduction factors (SRFs), since the expectation is that all welds are volumetrically examined. Therefore, the governing load combination is re-analyzed assuming the efficiency of the weld seam to be 80% (i.e., 0.8) of the design. This SRF was selected based on the known condition of the MPC shell-to-shell weld and consideration of Interim Staff Guidance ISG-15 incorporated into NUREG-2215; ASME Section VIII, Div. 1 Rules for Construction of Pressure Vessels; and ASME Section III, Rules for Construction of Nuclear Facility Components, Subsections ND, Class 3 Components, and Subsection NG, Core Support Structures.

The known condition of the shell-to-shell weld includes the weld type, the nature of the repair, and the inspections. The MPC longitudinal shell-to-shell seam weld is a full penetration butt weld that is classified as a Category A joint per NB-3351. After initial formation of the weld, the entire seam of the MPC shell-to-shell plate weld was volumetrically examined using radiography (RT). This original inspection revealed a lack-of-fusion indication that required local repairs between approximately 4 to 14 inches from the bottom of the MPC baseplate. Weld repairs were performed using a process that consisted of excavating the weld to remove the indication, performing a dye penetrant test (PT) examination to confirm removal of the indication, filling the excavated area with new weld material, and performing an RT examination of the completed weld repair. The first evacuated area was roughly 6.5 inches long, 5/8 inch wide and 9/32 inch in depth.

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-3 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 Subsequent to the first weld repair, another indication was found prompting a second weld repair and examination. This weld evacuation partially overlaid the first weld repair and was between approximately 8.5 and 25 inches from the bottom of the MPC base plate.

The second evacuated area was roughly 16.5 inches long, 5/8 inch wide and 1/4 inch in depth. As stated above, approximately 7.5 inches of the second weld repair does not have a documented RT examination. Final PT examinations were conducted of each weld repair with satisfactory results. The section of weld that was RT examined following the final weld repair did not have any further indications.

ASME Section III, Subsection NG, which applies to core support structures and has the same design stress intensity values as Subsection NB, specifies a quality factor of 0.75 for a Category A full penetration weld joint subjected to root and final PT only (i.e., no RT) per Table NG-3352-1. However, this is an overly conservative lower bound for the SRF associated with the shell-to-shell plate weld since more than 98% of the weld has been examined by RT.

ASME Section VIII, Div. 1, Table UW-12 and ASME Section III, Subsection ND-3352.1 for Category A joints both specify a SRF of 0.85 subject to spot radiography. As specified in UW-52 of Section VIII, Div. 1 and ND-5430 spot radiography requires a minimum of one 6-inch spot to be examined for every 50-foot increment of the weld. By comparison, more than 98% of the MPC welds have been examined by RT, far exceeding the requirement for spot radiography per Section VIII, Div. 1. Therefore, the 0.85 value for joint efficiency is considered conservative. It is also noted that, for MPC confinement boundary stainless steel material, the design stress values applicable to ASME Section VIII, Div. 1 and Section III, Subsection ND are generally equal to the design stress intensity values applicable to Section III, Subsection NB, except for minor variances at 300 and 400 degrees Fahrenheit.

The selected SRF of 0.8 for the re-evaluation of the structural loading conditions was based on consideration of the ASME codes previously discussed and NRCs code alternative for partial penetration lid-to-shell welds. This SRF introduces additional conservatism into the structural analysis and is based on the NRC position in ISG-15 now incorporated into NUREG-2215. This code alternative endorses a SRF of 0.8 for austenitic canisters with lid-to-shell (LTS) weld subject to progressive PT examination.

Although the weld in question is not a LTS weld, the weld repairs resulted in partial weld removal, PT examination to ensure the indication was removed, new weld material, and final PT examination of the repair.

Several conservatisms from the existing FSAR design basis analysis, such as using bounding pressures, temperatures, and temperature contours are maintained in the re-analysis. The structural re-analysis provided in Enclosure 3 shows that the factors of safety remain above 1.0, which demonstrates that the MPC can be accepted in the as-is state.

Confinement As described above, the MPC confinement boundary analyses demonstrate safety factors above 1.0, so the system maintains its confinement integrity under all normal, off-normal, and accident conditions. MPC 234 was subjected to helium leakage testing in accordance with CoC-1032 wherein the leaktight acceptance criteria is contained in ANSI N14.5-1997.

Holtec conducted leakage testing of the shell and lid during manufacturing and TVA Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-4 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 conducted testing of the closure welds during loading operations. The test results were satisfactory for each of the tests. Specific to the MPC shell test, that was conducted after the final weld repair, the measured leakage rate was 0.530 x 10 -8 std-cm3/sec against the leaktight criterion of 2.0 x 10 -7 std-cm-3/sec.

Criticality The above structural analysis shows that there is no change to the confinement of the system or to the basket within the MPC, and therefore no change to the criticality safety.

Shielding There is no change to the storage overpack or transfer cask, which provide the shielding for the fuel, and the same fuel is stored within the MPC in the same locations. Therefore, there is no change to the shielding safety of the system.

Thermal The MPC maintains its confinement integrity and helium environment, so there is no change to the thermal performance of the system.

Conclusion The above evaluation demonstrates that the MPC maintains all safety functions.

4. Basis for Approval of Exemption Request In accordance with 10 CFR 72.7, the Commission may, upon application by any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

a. Authorized by Law The Commission issued 10 CFR 72.7 under the authority granted to it under Section 133 of the Nuclear Waste Policy Act of 1982, as amended, 42 U.S.C. § 10153. Section 72.7 allows the NRC to grant exemptions from the requirements of 10 CFR Part 72. Granting the proposed exemption provides adequate protection to public health and safety, and the environment. As described below, the proposed exemption will not endanger life or property, or the common defense and security, and is otherwise in the public interest.

Therefore, the exemption is authorized by law.

b. Will not Endanger Life or Property or the Common Defense and Security TVA has provided an evaluation in the Justification Section above that shows the MPC structural protection and confinement boundary will continue to be maintained even with conservative limits applied to structural analyses. The storage systems criticality, shielding, and thermal analyses are not impacted by the unavailable RT examination.

Based on this, the proposed exemption does not endanger life or property or the common defense and security.

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-5 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3

c. Otherwise in the Public Interest The exemption will be in the public interest in that it will allow for continued interim storage of spent nuclear fuel at the SQN ISFSI without the need to defuel the storage canister to conduct an RT examination and maintain the result of that exam until termination of the license or CoC. Defueling the storage canister would result in additional handling of the storage system, creation of contaminated waste from removal of the MPC lid, and subject dried fuel to additional thermal cycles. The defueling and returning of the loaded MPC to storage will result in additional dose to plant personnel.

5. Environmental Consideration The potential environmental impact of using the Holtec HI-STORM FW MPC Storage System was analyzed in the environmental assessment for the final rule to add Amendment No. 3 to CoC-1032 to the list of approved spent fuel storage casks in 10 CFR 72.214 [Reference 2]. The environmental assessment for CoC-1032 Amendment No. 3 tiers off the July 18, 1990 [Reference 3] amendment to 10 CFR Part 72 to provide for the storage of spent fuel under a general license. NRC concluded the design of the cask would prevent the loss of confinement, shielding, and criticality control and thus result in insignificant environmental impacts. A determination was made that the amendment did not result in significant design or fabrication change of the cask; and there were no significant changes to cask design requirements in the proposed CoC amendment.

Therefore the occupational exposure or offsite dose rates from the amendment would remain well within the 10 CFR part 20 limits. It was also determined that changes to the CoC would not result in any radiological or non-radiological environmental impacts that significantly differ from the environmental impacts evaluated in the environmental assessment supporting the July 18, 1990, final rule. The NRC found there would be no significant change in the types or significant revisions in the amounts of any effluent released, no significant increase in the individual or cumulative radiation exposure, and no significant increase in the potential for or consequences from radiological accidents. NRC determined that an environmental impact statement was not necessary for the direct final rule of CoC-1032 Amendment No. 3.

The environmental impacts of the proposed exemption will not have an adverse impact to the environment, nor do they change assumptions in the previous environmental assessment. Therefore, the proposed action does not require any federal permits, licenses, approvals, or other entitlements.

a. Environmental Impacts of the Proposed Action The SQN ISFSI is located within the SQN site Protected Area. The ISFSI has a surrounding nuisance fence that is the current radiological control boundary. The area considered for potential environmental impact as a result of this exemption request is the area in and surrounding the ISFSI.

The interaction of loaded HI-STORM FW MPC Serial No. 234 with the environment is through the thermal, shielding, and confinement design functions for the DCS system.

This action has no impact on the MPC heat removal, the shielding capabilities of the DCS system, or the confinement boundary effectiveness.

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-6 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 There are no gaseous, liquid or solid effluents (radiological or non-radiological),

radiological exposures (worker or member of the public) or land disturbances associated with the proposed action.

Therefore, approval of the requested exemption to allow MPC Serial 234 to not have a complete volumetric examination record has no impact on the environment.

b. Adverse Environmental Effects Which Cannot be Avoided Should the Exemption be Approved As noted previously, there are no environmental impacts associated with approval of this exemption. Therefore, there are no adverse environmental effects which cannot be avoided should the exemption request be approved.

c. Alternative to the Proposed Action TVA considered an alternative to the proposed exemption request. The alternative would be to unload the MPC to perform the examination, reload the spent fuel, and place into storage in the ISFSI.

d. Environmental Effects of the Alternative to the Proposed Action TVA considered the alternative to the proposed exemption request on environmental impacts. The unload process as described in Section 9.4 of the HI-STORM FW FSAR, results in occupational dose accumulation and the creation of trash both contaminated and non-contaminated. Whether the same MPC is used to reload the fuel or a new MPC is used, occupational dose accumulation and the creation of radiological waste will occur. In summary of the unloading events: the HI-STORM cask will be retrieved from the ISFSI and brought into the 10 CFR Part 50 fuel storage structure; the MPC is lifted from its overpack into the transfer cask and then moved to a work stand location; the closure plates welds are removed to access the vent and drain ports; the MPC is flooded with borated water; the lid-to-shell weld is removed; the transfer cask is then placed in the spent fuel pool; and the spent fuel placed into storage racks. Although not an impact on the environment, unloading and loading the DCS system will entail financial expenditure and heavy load lift operations.

e. Conclusion and Status of Compliance TVA concludes that the proposed action, which will allow for the volumetric examination of the weld repair and its record to be omitted, is in the public interest in that it avoids the adverse environmental, radiological, and financial effects associated with the alternative to the proposed action.

Based on the above environmental considerations and justification information, TVA finds this action meets the criteria for categorical exclusion criteria of 10 CFR 51.22(c)(25) or otherwise does not require an environmental review, because there are:

(i) no significant hazards consideration; (ii) no significant change in the types or significant increase in the amounts of any effluents that may be releases offsite; (iii) no significant increase in individual or cumulative public or occupational radiation exposure; Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-7 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 (iv) no significant construction impact; (v) no significant increase in the potential for or consequences from radiological Accidents.

Finally, under 10 CFR 51.22(c)(25), the exemption being sought involves recordkeeping requirements and an inspection requirement covered by §51.22(c)(25)(vi)(A) and (C),

respectively

5. Precedent This exemption request is similar to the South Texas Project (STP) request recently approved by NRC in Reference 4. This exemption request is different in regard to the weld examination location, a shell-to-baseplate versus a shell-to-shell weld. TVA is also requesting exemption from 10 CFR 72.174, which is different from Reference 4.

6. References

1. Holtec International Final Safety Analysis Report on the HI-STORM FW MPC Storage System, Holtec Report No. HI-2114830, Revision 6, June 18, 2019

2. Federal Register, Vol. 82, No.123 Wednesday, June 28, 2017, pg 29225 List of Approved Spent Fuel Storage Casks: Holtec International HISTORM Flood/Wind Multipurpose Canister Storage System, Certificate of Compliance No. 1032, Amendment No. 3

3. Federal Register, Vol. 55, No.138 Wednesday, July 18, 1990, pg 29181, Storage of Spent Fuel in NRC-Approved Storage Casks at Power Reactor Sites

4. NRC letter dated April 25, 2022, Exemption from Certain Provisions of Title 10 of The Code of Federal Regulations, Sections 72.154, 72.212, And 72.214 HI-STORM Flood And Wind Multi-Purpose Canister Serial Number 248 Requirement For 100 Percent Inspection Of Shell-to-Baseplate Weld And Related Records Retention Requirements At South Texas Project Electric Generating Station Independent Spent Fuel Storage Installation (CAC No.001028; Docket Nos. 72-1041, 50-498, and 50-499; EPID: L-2022-LLE-0009)

(Agencywide Documents Access and Management System [ADAMS] Accession No. ML22094A176)

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 E1-8 of E1-8

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

INDEPENDENT SPENT FUEL STORAGE INSTALLATION Affidavit Pursuant to 10 CFR 2.390 to Withhold Information from Public Disclosure Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk TVA Letter - Sequoyah Nuclear Plant - Request for Exemption AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Kimberly Manzione, being duly sworn, depose and state as follows:

(1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is Enclosure 3 to the TVA Letter -

RRTI-2087-007 Rev 2, which contains Holtec Proprietary information.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk TVA Letter - Sequoyah Nuclear Plant - Request for Exemption AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a and 4.b above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to 2 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk TVA Letter - Sequoyah Nuclear Plant - Request for Exemption AFFIDAVIT PURSUANT TO 10 CFR 2.390 regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtecs competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk TVA Letter - Sequoyah Nuclear Plant - Request for Exemption AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

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Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3 ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

INDEPENDENT SPENT FUEL STORAGE INSTALLATION Response to Request for Technical Information RRTI-3087-007R2 (Proprietary)

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 3