ML23006A088
| ML23006A088 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/06/2023 |
| From: | John Lamb Plant Licensing Branch II |
| To: | Lowery K, Pournaras D Southern Nuclear Operating Co |
| References | |
| L-2022-LLA-0096 | |
| Download: ML23006A088 (11) | |
Text
From: John Lamb Sent: Friday, January 6, 2023 11:15 AM To: Lowery, Ken G.; Pournaras, DeLisa S.
Cc: Chamberlain, Amy Christine
Subject:
For Your Action - Request for Additional Information (RAI) - Vogtle - AST, TSTF-51, TSTF-471, and TSTF-490 LAR (L-2022-LLA-0096)
Importance: High DeLisa and Ken, By letter dated June 30, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession Package Number ML22181B066), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a License Amendment Request (LAR) to revise Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, Technical Specifications (TS). The proposed LAR would revise the Vogtle, Units 1 and 2, current licensing basis to implement an alternative radiological source term for evaluating design basis accidents as allowed by Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, Accident Source Term. In addition, the proposed LAR requests to incorporate Technical Specification Task Force (TSTF) Travelers TSTF-51-A, Revise containment requirements during handling irradiated fuel and core alterations, Revision 2; TSTF-471-A, Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes, Revision 1; and TSTF-490-A, Deletion of E Bar Definition and Revision to RCS Specific Activity Technical Specification, Revision 0.
The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the impact of implementing an alternative radiological source term for evaluating design basis accidents (DBAs) on all DBAs currently analyzed in the Vogtle, Units 1 and 2, updated final safety analysis report (UFSAR) that could have the potential for significant dose consequences. The NRC staff determined that more information is needed to complete its review.
On December 16, 2022, the NRC staff provided draft RAI questions to SNC to make sure that the RAIs were understandable, the regulatory basis is clear, to ensure there is no proprietary information, and to determine if the information was previously docketed. On January 6, 2023, NRC staff held a clarifying call with SNC and SNC stated that it would provide the RAI responses 31 days from the date of this email.
If you have any questions, you can contact me at 301-415-3100.
Sincerely, John REQUEST FOR ADDITIONAL INFORMATION (RAI)
By letter dated June 30, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession Package Number ML22181B066), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a License Amendment Request (LAR) to revise Vogtle
Electric Generating Plant (Vogtle), Units 1 and 2, Technical Specifications (TS). The proposed LAR would revise the Vogtle, Units 1 and 2, current licensing basis to implement an alternative radiological source term for evaluating design basis accidents as allowed by Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, Accident Source Term. In addition, the proposed LAR requests to incorporate Technical Specification Task Force (TSTF) Travelers TSTF-51-A, Revise containment requirements during handling irradiated fuel and core alterations, Revision 2; TSTF-471-A, Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes, Revision 1; and TSTF-490-A, Deletion of E Bar Definition and Revision to RCS Specific Activity Technical Specification, Revision 0.
The Nuclear Regulatory Commission (NRC) staff reviewed the impact of implementing an alternative radiological source term for evaluating design basis accidents (DBAs) on all DBAs currently analyzed in the Vogtle, Units 1 and 2, updated final safety analysis report (UFSAR) that could have the potential for significant dose consequences. The NRC staff determined that more information is needed to complete its review.
Regulatory Analysis Basis
- 1. Section 10 CFR Part 50.67, Accident Source Term, allows licensees seeking to revise their current accident source term in design basis radiological consequence analyses to apply for a license amendment under § 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report. Section 50.67(b)(2) requires that the licensee's analysis demonstrates with reasonable assurance that:
(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE.
(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.
- 2. Plant Design Criterion 19, Control Room, provides for conformance with Section 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, states, in part, that whenever a holder of an operating license under this part, desires to amend the license, application for an amendment must be filed with the Commission, as specified in § 50.4 of this chapter, as applicable, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.
- 3. Section 10 CFR Part 50, Appendix A, General Design Criterion (GDC) for Nuclear Power Plants: GDC 19, requires that a control room be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. It also states that adequate radiation protection is provided to permit access and occupancy of the control
room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
- 4. The regulation at Subpart H of 10 CFR 20, Standards for Protection against Radiation, provides the requirements for respiratory protections and controls to restrict internal exposure in restricted areas. Specifically, 10 CFR 20.1701 states that licenses shall use, to the extent practical, process or engineering controls to control the concentration of radioactivity in the air. Use of other controls as described in 10 CFR 20.1702 is allowed by regulation when it is not practical to apply process or other engineering controls to control the concentrations of radioactive material in the air.
- 5. Section 10 CFR 50.36, Technical Specifications, contains the NRCs regulatory requirements related to the content of the Technical Specifications (TSs). This regulation requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notifications; and (8) written reports.
- 6. NUREG-1431, Standard Technical Specifications Westinghouse Plants Revision 4.0, Volume 1, Specifications dated April 2012 contains the improved standard technical specifications (STS) for Westinghouse plants. The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 Federal Register (FR) 39132), which was subsequently codified by changes to 10 CFR 50.36 (60 FR 36953).
Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency.
- 7. NUREG-0800, Standard Review Plan (SRP) Section 6.4, Control Room Habitability System, Revision 3, March 2007; Section 6.5.2, Containment Spray as a Fission Product Cleanup System, Revision 4, March 2007; and Section 15.0.1, Radiological Consequences Analyses Using Alternative Source Terms, Revision 0, July 2000.
- 8. NRC Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ADAMS Accession No. ML003716792).
- 9. NRC RG 1.196, Control Room Habitability at Light-Water Nuclear Power Reactors, Revision 1, January 2007 (ADAMS Accession No. ML063560144).
- 10. NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, D.A.
Powers and S.B. Burson, USNRC, June 1993.
- 11. NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, D.A Powers, et.al., USNRC, July 1996.
RAI No. 1 (General)
Regulatory Basis numbered 1, 2, and 8 apply to RAI No. 1.
In the letter dated June 30, 2022, SNC states: For Vogtle Units 1 and 2, it is requested that 40% of the rods be allowed to exceed the 6.3kW/ft limit and those 40% of rods be approved for a LHGR limit of 7.4 kW/ft.
Does this mean 40% of the rods for a single assembly or 40% of the rods for all the assemblies?
RAI No. 2 (General)
Regulatory Basis for RAI No. 2 The regulation at 10 CFR 50.36(b) states, in part," the technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34." The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when a LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
Background
By letter dated October 4, 2018 (ADAMS Accession No. ML17346A587), the NRC staff informed the industry that after considerable review and analysis, the NRC staff concluded that for certain facilities, applications regarding adoption of TSTF-51 and TSTF-471 could result in exceeding the bounding licensing basis Fuel Handling Accident (FHA) analysis of record dose for the control room and is therefore considered an unanalyzed condition. Therefore, in order to support the NRC staffs review and approval of the licensees adoption of the these TSTFs, the NRCs letter suggested the licensees to include in their applications, certain information as specified in the letter.
RAI SNCs proposed change to Vogtle TS 3.9.4 Containment Penetrations, eliminates the use of the defined term CORE ALTERATIONS from the TS Applicability requirement per TSTF-51.
Since the proposed change involves a change to a Vogtle TS Applicability requirement, SNC is requested to address the NRC staffs letter dated October 4, 2018.
RAI No. 3 (General)
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 3.
On page 1 of the application, it states SNC requests Nuclear Regulatory Commission (NRC) review and approval of the proposed revisions to the licensing basis of VEGP [Vogtle Electric Generating Plant] that support a selected scope application of an Alternative Source Term (AST) methodology. In section 1.0 Summary Description it states Southern Nuclear Operating Company (SNC) requests Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of Vogtle that support a selective scope application of an Alternative Source Term (AST) methodology.
In all other sections the LAR is stated as a full scope implementation. (e.g. Table A:
Conformance with Regulatory Guide 1.183 Section C, Regulatory Position 1.1.3 it states Conforms - This is a full scope AST implementation for the radiological dose consequences of the VEGP Design Basis Accidents.)
Please address the discrepancy.
RAI No. 4 (Loss-of-Coolant Accident (LOCA))
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 4.
In the LAR, Attachment 4 contains the Conformance With Regulatory Guide (RG) 1.183 Appendix A (Loss of Coolant Accident), Attachment 5 to Enclosure, Loss-of-Coolant Accident Analysis lists the parameters and assumptions for the LOCA. Attachment 11 is the Vogtle AST Accident Analysis Input Values Comparison Tables and contains Table 2: LOCA Inputs and Assumptions. However, there are inconsistencies between these portions of the LAR Enclosures. , Response to Regulatory Position A-5.2 states:
- However, administrative limits ensure that the operational leakage outside of containment from Emergency Core Cooling System (ECCS) systems is limited to no more than 2 gallons per minute (gpm) total, which is multiplied by two.
- In addition, two times the assumed leak rate of 7.0 gpm passed valves that isolate return flow to the Refueling Water Storage Tank (RWST) is evaluated separately. and attachment 11 states:
- ECCS Leak Rate = 2.0 gal/min
- 1. Please provide clarification of the values in Table 2: LOCA Inputs and Assumptions for each of these parameters. It appears that your initial assumptions used for your dose calculations are using values of 2 gal/min for ESS Leak Rate, and 7gal/min for ECCS Leakage Rate to the RWST as stated in Attachment 11 Comparison Tables. This appears to be in conflict with the Conformance with Regulatory Guide tables.
- 2. Please update Table 2: LOCA Inputs and Assumptions to include AST values for ECCS Leakage to the RWST RAI No. 5 (LOCA)
Regulatory Basis numbered 1, 2, 4, 5, and 7 apply to RAI No. 5.
RG 1.183 Appendix A regulatory position 3.3 states:
Reduction in airborne radioactivity in the containment by containment spray systems that have been designed and are maintained in accordance with
Chapter 6.5.2, of the SRP [Standard Review Plan] (Ref. A-1) may be credited.
Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays1 (Ref. A-4). The simplified model is incorporated into the analysis code RADTRAD (Refs. A-1 to A-3).
In the LAR, Table B of Enclosure 4 states Vogtles conformance with RG 1.183 Appendix A.
Table B states that Vogtles analysis for RG 1.183 regulatory position 3.3 is, Conforms -
Containment Spray is credited for elemental and particulate iodine removal. Further in that section, it states that the containment spray elemental iodine removal coefficient is 13.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />s-1. 1 (page A11-3) provides comparison tables of the new AST values compared to the current licensing basis values. Table 2 of Attachment 11 states the LOCA inputs and assumptions; it states that the elemental iodine spray removal coefficient in the current licensing basis assumes 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s-1 and that the new AST assumes 13.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />s-1.
Please explain how the elemental iodine spray removal coefficient was calculated and discuss consistency with RG 1.183. Provide enough detail to allow the NRC staff to confirm the methodology is consistent with NUREG-0800 Chapter 6.5.2 and/or NUREG/CR-5966 as applicable.
Regulatory Basis numbered 1, 2, 7, 5, and 6 above apply to RAI No. 6.
Each accident analysis has a time for Control Room Isolation (CR Isolation) and Control Room Pressurization Mode Initiation (CR Pressurization Mode Initiation). The time difference between the two events varies with each of the accidents (e.g. 88 seconds for LOCA, 90 seconds for Fuel Handling Accident (FHA), and 80 seconds for Control Rod Ejection)
Please discuss the timing associated with the control room emergency filtration/pressurization system, include in the discussion at what time after the event a high radiation or Safety Injection Signal (depending on the event) will occur. Include in your response how long it takes the instrumentation to process the signal, how long it takes the control room ventilation system to reposition to the isolated position and/or pressurization mode. In addition, provide a comparison to the current licensing basis assumption.
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 7. , Table 1 lists the source term for the FHA. This list includes the following isotopes:
- Kr-83m, Kr-85
- Xe-131m, Xe-133, Xe-133m, Xe-135, Xe-135m, Xe-138
- Br-82, Br-83, Br-84
- I-130, I-131, I-132, I-133, I-134, I-135 , Table 1 lists the source term for the LOCA. This list includes the following isotopes:
- Kr-83m, Kr-85, Kr-85m, Kr-87, Kr-88
- Xe-131m, Xe-133, Xe-133m, Xe-135, Xe-135m, Xe-138
- Br-82, Br-83, Br-84
- I-130, I-131, I-132, I-133, I-134, I-135
- Cs-134, Cs-134m, Cs-135, Cs-136, Cs-137, Cs-138
- Rb-86, Rb,-88, Rb-89
- Sb-124, Sb-125, Sb-126, Sb-127, Sb-129
- Te-125m, Te-127, Te-127m, Te129, Te-129m, Te-131, Te-131m, Te-132, Te-133, Te-133m, Te-134
- Sr-89, Sr-90, Sr-91, Sr-92
- Ba-137m, Ba-139, Ba-140 RG 1.183 Appendix B regulatory position 1.2 states:
The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.
All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.
The source term provided in Tables 1 in the FHA analysis deviates from RG 1.183 Appendix B regulatory position 1.2 and excludes radionuclides. Please explain the deviation from the RG 1.183 or conform to RG 1.183.
RAI No. 8 (Main Steam Line Break (MSLB))
Regulatory Basis number 5 applies to RAI No. 8. to Enclosure 1, Main Steam Line Break Analysis Contains Table 4 - MSLB Flow Rates. The flow path Faulted SG to Env has a flow of 1000 cubic feet per minute (cfm) for 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (hr). The total Release (pounds mass (lbm)) contains no values.
Please update the Table.
RAI No. 9 (Control Rod Ejection Accident (CREA))
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 9.
In Attachment 4, Table G: Conformance with Regulatory Guide 1.183, Appendix H (Rod Ejection Accident, Regulatory Positions H-5 to H-7.4 are missing.
Please provide updated table.
RAI No. 10 (CREA)
Regulatory Basis numbered 1, 2, 7, and 5 apply to RAI No. 10.
In the LAR, Section 3.8 of Enclosure 1 states that for the CREA that No credit is taken for removal by containment sprays or for deposition of elemental iodine on containment surfaces 1 provides comparison tables of the new AST values compared to the current
licensing basis values. Table 6 states the CREA inputs and assumptions; it states that for natural deposition in containment the current licensing basis assumes 50% plateout of the reactor coolant system release and that the new AST assumes an aerosol removal rate of 3.005E-2 hr-1 and no removal of elemental iodine. Table 6 states that the reason for the change is natural deposition is credited per RG 1.183 Appendix H Section 6.1.
RG 1.183 Appendix H regulatory position 6.1 states that a reduction in the amount of radioactive material available for leakage from containment that is due to natural deposition may be taken into account and it refers to RG 1.183 Appendix A for guidance on acceptable methods and assumptions for evaluating this mechanism. RG 1.183 Appendix A states that reduction in airborne radioactivity in the containment by natural deposition within the containment may be credited and that acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2 of NUREG-0800 and in NUREG/CR-6189.
Table 36 of NUREG/CR-6189 lists five specific time intervals and their correlations. However, lists one value for natural deposition and does not provide enough information for the NRC staff to determine that this value reflects a methodology consistent with NUREG/CR-6189.
Please provide a summary of the methodology used in enough detail to allow the NRC staff to determine consistency with NUREG/CR-6189.
RAI No. 11 (CREA)
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 11.
RG 1.183 Appendix H, Assumptions for Evaluating the Radiological Consequences of a pressurized water reactor (PWR) Rod Ejection Accident, regulatory position 4 states:
The chemical form of radioiodine released to the containment atmosphere should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15%
organic iodide. If containment sprays do not actuate or are terminated prior to accumulating sump water, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.
In the LAR, Table G in Attachment 4, Section H-4, states Vogtles conformance with RG 1.183 Appendix H. Table G states that Vogtles analysis for RG 1.183 regulatory position 4 is:
Conforms - The chemical form of radioiodine released to the containment atmosphere is assumed to be 95% cesium iodide, 4.85% elemental iodine, and 0.15% organic iodide. Since containment sprays are not assumed to be activated in this event, no credit is taken for pH being controlled at values of 7 or greater.
Because containment sprays are not actuated and no credit is taken for the containment sump pH being controlled at values of 7 or greater, the iodine species needs to be evaluated on a plant specific basis to determine that 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide is a conservative assumption.
Provide the plant specific evaluation that determined that the chemical form of radioiodine released to the containment atmosphere of 95% Csl, 4.85% elemental iodine, and 0.15%
organic iodide is conservative at Vogtle and that the iodine does not re-evolve. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g.,
pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.
Regulatory Basis numbered 1 applies to RAI No. 12.
The NRC staff has verified the value of 659 micro Curie per gram (µCi/gram) using the Nominal reactor coolant system (RCS) Noble Gas concentrations based on 1% cladding defects as listed in a Table staring on page E-24 of the LAR. The NRC staff notes that there appears to be some minor discrepancies between the values in the Nominal Noble Gas concentrations listed in the LAR and the values in Vogtle FSAR Table 11.1-2, Reactor Coolant Design Basis Fission and Corrosion Product Specific Activity.
Please provide additional information describing the derivation of the proposed Dose Equivalent Xenon 133 (DE Xe-133) limit of 280 µCi/gram.
RAI No. 13 (CREA and Locked Rotor Accident (LRA))
Regulatory Basis numbered 1, 2, and 5 apply to RAI No. 13.
In Section 3.5, Fuel Handling Accident, it is stated:
The FHA analysis used assumptions and inputs that follow the guidance in RG 1.183.
The key parameters and assumptions are listed in Attachment 6. An exception to the RG-1.183 linear heat generation limit (LHGR) of 6.3 kW/ft (footnote 11) has been requested. For Vogtle Units 1 and 2, it is requested that 40% of the rods be allowed to exceed the 6.3kW/ft limit and those 40% of rods be approved for a LHGR limit of 7.4 kW/ft.
In Section 3.2.1, Fission Product Inventory, it is stated:
The nominal inventory of fission products in the reactor core was calculated using ORIGEN-ARP based on the full power operation of the core plus uncertainty. The nominal inventory was based on an equilibrium cycle modeled with lead rod burnup of 62 [giga Watt day per metric ton of uranium] GWD/MTU and variable enrichment regions.
In Attachment 4, Table A: Conformance with Regulatory Guide 1.183 Section C, Regulatory Position 3.2 states:
For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.
The Vogtle Analysis states that it conforms to the Regulatory Position, and thereby conforms with footnote 11 associated with Table 3.
The key assumptions of rod burnup of 62 GWD/MTU and LHGR of 7.4 kW/ft is not addressed in analysis of any other Design Basis Accident where there is fuel damage.
Please provide additional information on how the key assumptions associated with burnup and LHGR (where they exceed the values in footnote 11) are not applicable LRA and CRDA which assumes fuel damage.
Hearing Identifier: NRR_DRMA Email Number: 1890 Mail Envelope Properties (MN2PR09MB50846C2EFF6010E6D1D118BBFAFB9)
Subject:
For Your Action - Request for Additional Information (RAI) - Vogtle - AST, TSTF-51, TSTF-471, and TSTF-490 LAR (L-2022-LLA-0096)
Sent Date: 1/6/2023 11:14:32 AM Received Date: 1/6/2023 11:14:00 AM From: John Lamb Created By: John.Lamb@nrc.gov Recipients:
"Chamberlain, Amy Christine" <ACCHAMBE@southernco.com>
Tracking Status: None "Lowery, Ken G." <KGLOWERY@southernco.com>
Tracking Status: None "Pournaras, DeLisa S." <DSPOURNA@SOUTHERNCO.COM>
Tracking Status: None Post Office: MN2PR09MB5084.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 25337 1/6/2023 11:14:00 AM Options Priority: High Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: