NL-23-0010, Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490

Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490

ML23037A856
Person / Time
Site:	Vogtle
Issue date:	02/06/2023
From:	Gayheart C Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-23-0010
Download: ML23037A856 (1)

Text

Cheryl A. Gayheart 3535 Colonnade Parkway Regulatory Affairs Director Birmingham, AL 35243 205 992 5316 cagayhea@southernco.com February 6, 2023 NL-23-0010 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 and 2 Docket Nos. 50-424 & 50-425

Subject:

Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 By letter dated June 30, 2022, Southern Nuclear Operating Company (SNC) submitted an application to 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 requested 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.

By email dated January 6, 2023, the Nuclear Regulatory Commission (NRC) staff issued a request for additional information (RAI). The Enclosure to this letter provides the SNC response to the NRC staffs RAIs.

The conclusions of the No Significant Hazards Consideration and Environmental Consideration contained in the original application have been reviewed and are unaffected by this response.

U. S. Nuclear Regulatory Commission NL-23-0010 Page 2 If you have any questions, please contact Amy Chamberlain at 205.992.6361.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 6th day of February 2023.

Respectfully submitted, Cheryl A. Gayheart Director, Regulatory Affairs Southern Nuclear Operating Company CAG/kgl/cg

Enclosure:

SNC Response to NRC RAIs cc: Regional Administrator, Region ll NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector - Vogtle 1 & 2 State of Georgia Environmental Protection Division RType: CVC7000

Vogtle Electric Generating Plant - Units 1 and 2 Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Enclosure SNC Response to NRC RAIs

Enclosure to NL-23-0010 SNC Response to NRC RAIs NRC RAI-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?

SNC Response to NRC RAI-1:

No more than 40% of the fuel in any given fuel assembly will exceed the RG 1.183, Revision 0, Table 3, Footnote 11 gap fraction applicability limits (peak burnup up to 62,000 MWD/MTU and a maximum of 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU). There is no limit on the number of assemblies that this applies to.

NRC RAI-2:

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.

SNC Response to NRC RAI-2:

NRC letter ML17346A587 specifies regulatory requirements for plants implementing TSTF-51 and TSTF-471.

In the above-mentioned letter are several requirements that need to be satisfied:

1. Show that the Fuel Handling Accident (FHA) analysis can meet applicable dose acceptance criteria specified in 10 CFR 50.67 (relevant regulation for plants using an Alternate Source Term (AST)) only crediting systems that are required to be operable.

2. Licensees whose licensing basis includes analysis for the dropping of a heavy load onto irradiated fuel in Ch. 15 (or equivalent) of their updated Final Safety Analysis Report (FSAR) must analyze dropping of a heavy load as a part of adopting TSTF-51.

SNCs TSTF-51 and TSTF-471 submittal complies with items 1 and 2 above based on the following:

A. The FHA analysis described in SNCs submittal (Enclosure Section 3.5) only takes credit for the control room emergency filtration system (CREFS) to mitigate the FHA event, as CREFS is required to be operable during movement of recently irradiated fuel per TS 3.3.7. The FHA analysis described in Section 3.5 justifies that a release from the fuel handling building bounds the release from containment due to an open equipment hatch or personnel airlock (PAL). This justification does not take credit for the containment purge system being in operation, but rather shows that the dispersion factors for a release from the fuel building are bounding compared to a direct release from containment in either an open or closed configuration. Enclosure, Section 3.14 of SNCs submittal clarifies that E-1

Enclosure to NL-23-0010 SNC Response to NRC RAIs since CREFS is credited for FHA, a variation to the NUREG-1431 markups in TSTF-51 is taken for TS 3.3.7, and no changes are made to the TS. Table 3.5b of the Enclosure of SNCs submittal confirms the FHA event meets all acceptance criteria for RG 1.183 Rev.

0, 10 CFR 50.67 and GDC 19.

B. The Vogtle Units 1 and 2 FSAR Section 9.1.5.3.1.1.3 discusses the controls in place to prevent the drop of a heavy load onto exposed irradiated assemblies. Due to identification and use of safe load paths, the only heavy load that is available to be dropped on irradiated assemblies is that of the polar crane main hoist load block. Section 9.1.5.3.1.1.3 discusses that administrative controls are in place which prevent the main hoist from functioning during movement over exposed irradiated fuel. This eliminates the potential for a load drop that could result in damage to irradiated fuel assemblies in the core.

SNCs response A and B satisfy requirements 1 and 2 identified above. In addition to the above requirements for TSTF-51, ML17346A587 requires one of four bulleted statements to be addressed as a part of adopting TSTF-51 and TSTF-471. SNCs response will address the third bullet:

Describe the limitations or controls that would prevent movement of any unirradiated fuel assembly, source, reactivity control component, or other component affecting reactivity within the reactor vessel capable of damaging a fuel assembly prior to the time period defined as recently SNCs Refueling Operations (Mode 5 to Mode 6) procedure (12007-C) provides limitations and controls on core alterations (movement of recently irradiated fuel) within the reactor vessel. This procedure also satisfies Plant Vogtles Technical Requirements Manual 13.9.1 Decay Time. This procedure:

Prevents core alterations (movement of recently irradiated fuel) in the reactor vessel prior to the decay time analyzed in the FHA event. This corresponds to 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> for SNCs submittal.

Requires containment equipment hatch and personnel airlock to either be closed or be able to be closed by designated personnel in a timely manner.

Requires confirmation that containment penetrations or other access pathways are confirmed or capable of being secured.

In conclusion, the FHA analysis contained in SNCs AST submittal demonstrates all acceptance criteria are met as specified in RG 1.183 Rev. 0, 10 CFR 50.67, and GDC 19. The FHA analysis does not take credit for any system not required to be operable per Plant Vogtle Technical Specifications. The potential for a drop of a heavy load on exposed irradiated fuel assemblies is administratively controlled as discussed in Vogtle FSAR section 9.1.5.3.1.1.3. Vogtle plant procedures confirm containment penetrations or other access pathways are secured or capable of being secured in a timely manner prior to the start of core alterations (or movement of recently irradiated fuel), and that core alterations (movement of recently irradiated fuel) cannot begin prior to the decay time of 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> analyzed in the FHA event. As a result, SNCs TSTF-51 and TSTF-471 submittal meets the requirements specified in ML17346A587.

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Enclosure to NL-23-0010 SNC Response to NRC RAIs NRC RAI-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.

SNC Response to NRC RAI-3:

The Alternative Source Term (AST) implementation at Plant Vogtle Units 1 and 2 is a full scope implementation as defined by RG 1.183 Section C, Regulatory Position 1.2.1.

Markups to clarify the correct terminology on page 1 and Section 1.0 of the submittal are included below:

Page 1 Correction:

SNC requests Nuclear Regulatory Commission (NRC) review and approval of the proposed revisions to the licensing basis of VEGP that support a full selected scope application of an Alternative Source Term (AST) methodology.

Page E-3 Correction:

Southern Nuclear Operating Company (SNC) requests Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of VEGP that support a full selective scope application of an Alternative Source Term (AST) methodology.

E-3

Enclosure to NL-23-0010 SNC Response to NRC RAIs NRC RAI-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

• ECCS Leakage Rate to the RWST = 7.0 gal/min Please provide:

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 SNC Response to NRC RAI-4:

1. Attachment 5 of the License Amendment Request (LAR) provides the key assumptions and inputs for the Loss-of-Coolant-Accident (LOCA) dose consequences analysis including the Emergency Core Cooling System (ECCS) leakage and Refueling Water Storage Tank (RWST) leakage parameters. The parameter for a leakage rate of 2.0 gallons per minute for ECCS leakage that is provided represents the operational leakage limit applicable to Plant Vogtle Units 1 and 2. Similarly, the ECCS leakage rate to the RWSTs operational leakage limit of 7.0 gallons per minute is provided. The operational limits are provided as the input into the LOCA dose consequence analysis where the analysis methodology dictates that they are increased by a factor of two consistent with the methodology described by Regulatory Guide 1.183, Appendix A. The AST LOCA Dose Analysis describes the modeling of the analytical ECCS leakage where it defines the leak rate to be twice the design input (operational) leak rate of 2 gallons per minute and that 4 gallons per minute value is what is used to calculate dose. Similarly, the AST LOCA Dose Analysis describes that the design input for the RWST operational E-4

Enclosure to NL-23-0010 SNC Response to NRC RAIs backleakage of 7 gallons per minute is doubled to 14 gallons per minute which is used in the dose analysis.

Table 2 in Attachment 11 of the LAR is intended to compare the input values used in the current licensing basis LOCA dose consequence evaluation to that proposed by the new Alternative Source Term LOCA dose consequence evaluation. Due to the current approved licensing basis methodology not considering ECCS leakage back to the RWST, there is not a value in the current basis to compare to. Therefore, that value is not included in the table. Table 2 does provide the comparison of the ECCS leakage flow rate and similar to Attachment 5 of the LAR, it provides the operational ECCS leakage limit which is the input into the analysis.

The information in Attachment 4 of the LAR regarding conformance to Regulatory Guide 1.183, Regulatory Position A-5.2 is correct as it is describing the doubling of the operational leakage limits, which are the inputs into the analysis, to determine the analytical leakage limits for ESF leakage used in the dose consequence analysis.

2. Table 2 of Attachment 11 to the License Amendment Request is intended to compare the input values used in the current licensing basis LOCA dose consequence evaluation to that proposed by the new Alternative Source Term LOCA dose consequence evaluation.

Due to the current approved licensing basis methodology not specifically considering ECCS leakage back to the RWST, there is not a value in the current basis to compare to.

Therefore, that value was not included in the table.

To illustrate the changes discussed in the text above, a revised Attachment 11 Table 2 is provided below (additions are in blue and underlined):

Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Reason for Change Control Room Offsite and Control Room Containment Purge Iodine Chemical Form 5% particulate, 91% 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

elemental, 4% organic elemental, 0.15% organic Containment Volume 2,930,000 ft3 2,930,000 ft3 No change Containment Purge 0% 0% No change Filtration Removal by Wall None None No change Deposition Removal by Sprays None None No change Containment Leakage Iodine Chemical Form 5% particulate, 91% 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

elemental, 4% organic elemental, 0.15% organic Containment Sump >7.0 >7.0 No change pH Containment Sprayed 2,300,000 ft3 2,300,000 ft3 No change Volume Containment 630,000 ft3 630,000 ft3 No change unsprayed Volume Containment Spray 0 seconds 110 seconds Provides additional conservatism to Start Time Containment Leakage Pathway.

Containment Spray 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> No change Stop Time Containment Spray 2500 gpm 2500 gpm No change Flow Rate Elemental Iodine 10 hr-1 13.7 hr-1 Revision is consistent with RG 1.183 Spray Removal Appendix A RP 3.3 Coefficient E-5

Enclosure to NL-23-0010 SNC Response to NRC RAIs Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Reason for Change Control Room Offsite and Control Room Aerosol Spray 4.2 hr-1 5.34 hr-1 Revision is consistent with RG 1.183 Removal Coefficient Appendix A RP 3.3 Organic Iodine Spray None None No change Removal Natural Deposition Elemental, Organic- None, Elemental, Organic iodine Aerosol natural deposition is permitted per Aerosol - None - None, Aerosols - 0.1 hr-1 Appendix A of RG 1.183.

Containment Leakage Additional 5% conservative margin applied Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.2%/day 0.21%/day 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 30 days 0.1%/day 0.105%/day Containment Leakage 0% 0% No change Filtration ECCS Leakage to the Auxiliary Building Iodine Chemical Form 5% aerosol, 91% elemental, 0% aerosol, 97% The revised percentages are as specified in 4% organic elemental, 3% organic RG 1.183.

Containment Sump 115,000 ft3 114,922 ft3 Lower value without rounding is more Volume conservative due to higher sump concentrations ECCS Recirculation 30 minutes 30 minutes No change Start Time ECCS Leakage Flow 2 gpm/2 gpm 2 gpm/4 gpm The revised analytical value for ECCS Rate (input/analysis leakage is twice the operational leakage value) limit in accordance with RG 1.183 Appendix A RP 5.2.

ECCS Flashing 10% 10%

Fraction ECCS Leakage to the Not Modeled 7 gpm/14 gpm Consideration of ECCS leakage back to the RWST (input/analysis RWST is consistent with RG 1.183 value) Appendix A RP 5.2 NRC RAI-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. Attachment 11 (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 E-6

Enclosure to NL-23-0010 SNC Response to NRC RAIs methodology is consistent with NUREG-0800 Chapter 6.5.2 and/or NUREG/CR-5966 as applicable.

SNC Response to NRC RAI-5:

Regulatory Guide 1.183 Rev. 0 Section 3.3 specifies NUREG-0800 Chapter 6.5.2 is an acceptable model for iodine removal from sprays.

The elemental iodine spray removal coefficient of 13.7 hr-1 is a historical value provided by Westinghouse in Reference 1. It is based upon a reference plant and was developed using a method which pre-dated Revision 2 of NUREG-0800 Chapter 6.5.2.

While the details of the method used to generate the elemental iodine spray removal coefficient are not specified in Reference 1, the value of 13.7 hr-1 is conservative with respect to the Vogtle specific coefficient which is calculated based upon Revision 4 to NUREG-0800 Chapter 6.5.2 as illustrated below.

Section III.4.C.i of NUREG-0800 Chapter 6.5.2 provides the following relationship for the spray removal coefficient for elemental iodine:

6 Where:

s = Elemental iodine removal coefficient Kg = Gas-phase mass-transfer coefficient T = Spray drop fall time F = Spray volumetric flow rate V = Sprayed compartment volume D = Mass mean spray drop diameter For Vogtle Units 1 and 2, the plant specific spray removal coefficient is calculated (see below) to be s = 22.58 hr-1 using the above NUREG-0800 relationship.

6 6 689 0.00324 20100 22.58 2930000 0.00407 Where:

Kg = 689 ft/hr T = 3.24E-3 hr F = 2500 gpm = 2.01E4 ft3/hr V = 2.93E6 ft3 D = 0.124 cm = 4.07E-3 ft E-7

Enclosure to NL-23-0010 SNC Response to NRC RAIs It should be noted that Section III.4.C.i of NUREG-0800 Chapter 6.5.2 requires that s be limited to no more than 20 hr-1. Therefore, a value of 20 hr-1 would be applicable to the Vogtle containment using the NUREG-0800 methodology and confirms the value of 13.7 hr-1 used in the analysis is conservative.

References

1. WCAP-11611, Methodology for Elimination of the Containment Spray Additive, March 1988 NRC RAI-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.

SNC Response to NRC RAI-6: to Enclosure, Control Rod Ejection Accident Analysis, (page A9-9) lists an incorrect value for the CR Pressurization Mode Initiation time. The correct value used in the CREA analysis is 219 seconds (vs. 211 seconds given in Attachment 9 of the LAR), which represents a time difference of 88 seconds between CR Isolation (CRI) and CR Pressurization (versus the 80 seconds value indicated in the RAI), consistent with the LOCA analysis. A corrected table for to Enclosure (page A9-9) is provided below.

Control Room Ventilation Parameters Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 131 Seconds CR Pressurization Mode Initiation Automatic at 219 211 Seconds CR Ventilation System Normal Flow Rate 2,575 cfm < 131 seconds CR Ventilation System Filtered Makeup Rate 1,800 cfm > 211 seconds CR Ventilation System Recirculation Flow Rate 31,000 cfm > 211 seconds CR Ventilation System Charcoal Filter Efficiencies (Supply and Recirculation use the same filter)

All Iodine Species 99%

Particulates 99%

CR Unfiltered In-leakage 180 cfm > 131 seconds CR Ingress/Egress Unfiltered In-leakage 10 cfm > 0 seconds CR Breathing Rate 3.5E-4 m3/sec E-8

Enclosure to NL-23-0010 SNC Response to NRC RAIs Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 A CRI signal is generated by either a Safety Injection Signal (SIS) or by one of the in-vent radiation monitors in the CR intakes (Hi Rad). The time to generate an SIS or Hi Rad signal is event specific.

This information is provided for the AST LOCA, FHA, and Control Rod Ejection Accident (CREA) analyses in the following table.

Time Event Description

[sec]

SIS setpoint reached at 1.77 sec + 1.5 sec signal LOCA 3.3 delay = 3.27 sec (rounded up)

Bounding value selected based on 543 sec to reach FHA 600 the Hi Rad setpoint (including detector response time)

Time to reach SIS setpoint (121 sec) based on a 2 in2 CREA-induced small break LOCA and available CREA 123 timing for similar size LOCAs + 1.5 sec signal delay

= 122.5 sec (rounded up)

The LOCA time to generate an SIS of 3.3 seconds is consistent with the current licensing basis (CLB). The CLB FHA assumes that a CR Hi Rad signal is generated in 40 seconds, compared to the bounding value of 600 seconds used in the AST analysis. Control Room doses are only calculated for the LOCA and FHA in the CLB.

The time from SIS or Hi Rad signal generation to achieve CRI and CR Pressurization is generally not event-specific and is as given in the following table. The 8 seconds to achieve CRI includes 2 seconds of signal delay and 6 seconds for damper re-positioning. The 96 seconds from SIS /

Hi Rad signal generated to CR Pressurization mode achieved includes 36 seconds from SIS / Hi Rad generated to lead fan start signal + 30 seconds from lead fan start signal to system low flow detection + 30 seconds from lead fan low flow detection / lag fan start signal to CR pressure established (i.e., all events implicitly assume failure of lead fan to pressurize the CR).

Isolation Pressurization Mode [sec] Mode [sec]

Total time delay from SIS / Hi Rad generated to Mode 8.0 96 achieved The CRI and CR Pressurization timing shown is consistent with the CLB LOCA. The CLB FHA conservatively assumes an additional 2 seconds for Pressurization mode to be achieved (total of 98 seconds from Hi Rad signal). The AST FHA analysis also includes the additional 2 seconds, consistent with the CLB. This is the reason for the difference between the 88 seconds from CRI to CR Pressurization used in the LOCA and CREA analyses and the 90 seconds used in the FHA analysis.

The total time from event initiation to CRI and CR Pressurization used in the AST LOCA, FHA, and CREA analyses are shown in the following table.

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Enclosure to NL-23-0010 SNC Response to NRC RAIs Time from Event Initiation Time from Event Initiation Event to CR Isolation [sec] to CR Pressurization [sec]

LOCA 11.3 99.3 FHA 608 698 CREA 131 219 The CRI and CR Pressurization times for the AST LOCA and FHA analyses are compared to the CLB times in the following table.

Time from Event Initiation Time from Event Initiation Event to CR Isolation [sec] to CR Pressurization [sec]

AST LOCA 11.3 99.3 CLB LOCA 11.3 99.3 AST FHA 608 698 CLB FHA 48 138 NRC RAI-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.

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Enclosure to NL-23-0010 SNC Response to NRC RAIs 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.

SNC Response to NRC RAI-7:

SNCs fuel handling accident (FHA) analysis complies with RG 1.183.

The RG 1.183 Appendix B Regulatory Position 1.2 identifies radionuclides (xenons, kryptons, halogens, cesiums, and rubidiums) that should be considered for the gap activity. The verb consider is defined as to think carefully about, especially in order to make a decision.

All radionuclides listed in RG 1.183 Section C Regulatory Position 3.4, including those above, were considered. Those that would not contribute to the dose consequences were identified and eliminated from further evaluation. This was determined via a one-step screening process.

The screening process was based on the elements state. Per RG 1.183 Appendix B Regulatory Position 3, particulates, (i.e., solids), will be retained by the water in the spent fuel pool or refueling cavity. This is because Vogtle Technical Specifications 3.7.15 and 3.9.7 require a minimum level of 23 feet above the top of active fuel in the refueling cavity and spent fuel pool. As a result, only the noble gases (xenons and kryptons) and halogens (iodine and bromine) screen in as a part of the accident specific FHA source term.

Group Elements State Noble Gases Xenon and Krypton Gas Halogens Iodine and Bromine Gas Alkali Metals Cesium and Rubidium Solid Tellurium Group Tellurium, Antimony, Selenium, Barium, and Solid Strontium Noble Metals Ruthenium, Rhodium, Palladium, Molybdenum, Solid Technetium, and Cobalt Lanthanides Lanthanum, Zirconium, Neodymium, Europium, Solid Niobium, Promethium, Praseodymium, Samarium, Yttrium, Curium, and Americium Cerium Cerium, Plutonium, and Neptunium Solid This screening process resulted in the following radionuclides selected for determining the fuel handling accident dose consequences:

Halogens: I-130, I-131, I-132, I-133, I-134, I-135, Br-82, Br-83, Br-84 Noble Gases: Kr-83m, Kr-85, Kr-85m, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, and Xe-135, Xe-135m, Xe-138 The above nuclides are consistent with that listed in Attachment 6, Table 1 of the LAR.

E-11

Enclosure to NL-23-0010 SNC Response to NRC RAIs NRC RAI-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.

SNC Response to NRC RAI-8:

In the event of a Main Steam Line Break (MSLB), the complete contents of the secondary side of the faulted steam generator will blow down to the environment. This will occur very quickly upon accident initiation. Due to the way RADTRAD handles activity holdup and release from compartments, the explicit mass released from the faulted generator is not meaningful and is not used in the accident dose analysis. A conservatively large volumetric release rate is chosen, in this case 1000 cfm, which will allow for all the initial activity on the secondary side of the faulted steam generator to be released within a half hour. Attachment 7 to Enclosure, Table 4, Note 3 (page A7-5) provides an explanation of the use of 1000 cfm. This is consistent with the current licensing basis MSLB analysis for Vogtle Units 1 and 2.

NRC RAI-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.

SNC Response to NRC RAI-9:

The entire RG 1.183 Appendix H conformance table (updated with Regulatory Positions H-5 to H-7.4) is provided in the Attachment to this Enclosure. The conformance statement for RG Position H-4 has been updated consistent with the response to RAI No. 11.

NRC RAI-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 Enclosure 11 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 E-12

Enclosure to NL-23-0010 SNC Response to NRC RAIs 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.

SNC Response to NRC RAI-10:

The AST CREA analysis calculates an aerosol deposition removal coefficient (hr-1) using the 10th percentile Powers natural deposition model correlations for each of the five (5) time periods, using the six (6) correlations, given in NUREG/CR-6189, Table 36. The removal coefficients are calculated using a thermal power (P) of 3636 MWt.

The phase, time interval, correlation, and resulting calculated removal coefficient for each interval from NUREG/CR-6189 Table 36 (NUREG/CR-6604 Table 2.2.2.1-1) are given in the table below.

The analysis selects the bounding (smallest) removal coefficient (Gap phase from 0 - 0.5 hr) and conservatively applies it across all the time intervals. The magnitude of conservatism in this approach is evident from inspecting the range of removal coefficients calculated.

Powers 10th Percentile Natural Deposition for CREA Release Time Correlation Removal Phase Interval (hr) Coefficient (hr-1)

Gap 0 - 0.5 0.0182 + 3.26E-06

• P 3.005E-02 Gap 0.5 - 1.8 0.0645 * (1 - exp(-0.938*P/1000)) 6.237E-02 Early in-0.5 - 1.8 0.0326 * (1 - exp(-0.910*P/1000)) 3.141E-02 vessel Gap + early 1.8 - 3.8 0.094 * (1 - exp(-0.869*P/1000)) 9.001E-02 in-vessel Gap + early 3.8 - 13.8 0.0811 + 10.15E-06

• P 1.180E-01 in-vessel Gap + early 13.8 - 22.22 0.086 * (1 - exp(-2.384*P/1000)) 8.599E-02 in-vessel NRC RAI-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 E-13

Enclosure to NL-23-0010 SNC Response to NRC RAIs 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.

SNC Response to NRC RAI-11:

During the CREA, pressure in containment is not expected to reach the containment spray actuation set point. However, the Plant Vogtle containment sump houses baskets of trisodium phosphate (TSP).

Plant Vogtle Technical Specification 3.5.6 controls the amount of TSP. SNC assumes the event introducing water into the sump causes dissolution of TSP, thereby maintaining the pH of the sump water at 7 or greater. Given that the pH condition of RG 1.183 is met, it may be assumed that iodine in the containment atmosphere is 95% particulate, 4.85% elemental, and 0.15% organic.

In conclusion, the sump pH is controlled above 7 using large quantities of TSP additive. As a result, sensitivities on Iodine composition resulting from re-evolution are not required and the chemical form of Iodine specified in Regulatory Guide 1.183 Section C, Regulatory Position 3.5 and Appendix H, Regulatory Position 4 is appropriate and conservative. , Table G: Conformance with Regulatory Guide 1.183, Appendix H (Rod Ejection Accident), Regulatory Position H-4 has been updated accordingly. (See SNC Response to NRC RAI-9 and Attachment to this Enclosure)

NRC RAI-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 starting on page E-24 of the LAR. The NRC staff notes that there appears to be some E-14

Enclosure to NL-23-0010 SNC Response to NRC RAIs 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.

SNC Response to NRC RAI-12:

The differences between the nominal Noble Gas concentrations listed in FSAR Table 11.1-2 are due to updated coolant source terms developed in support of the Vogtle AST project that utilized the ORIGEN-ARP tool. These updated noble gas concentrations were communicated to the staff in SNCs submittal (pages E-24 and E-25).

If TSTF-490 were being implemented alone, the proposed plant-specific dose equivalent (DE)

Xe-133 Technical Specification limit would be selected to avoid impact to existing dose analyses (i.e., ~659 µCi/gram DE Xe-133). However, SNC is revising all dose analyses to support the AST LAR. Therefore, SNC has elected to specify a more restrictive DE Xe-133 Technical Specification limit.

The proposed Technical Specification limit of 280 Ci/g DE Xe-133 provides sufficient margin to dose acceptance criteria for the dose analyses performed in accordance with Reg Guide 1.183 Revision 0. Additionally, the proposed Tech Spec limit is two (2) to three (3) orders of magnitude greater than the DE Xe-133 concentrations measured in the previous five years of plant data from Vogtle 1 and 2.

NRC RAI-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 E-15

Enclosure to NL-23-0010 SNC Response to NRC RAIs 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.

SNC Response to NRC RAI-13:

The RG 1.183, Revision 0, Table 3 gap fractions are applicable to all non-LOCA events that consider failed fuel, other than Fuel Handling Accident (FHA), which considers gap fractions based on Table 2.9 of PNNL-18212, Rev. 1. Note that Attachment 4, Table A: Conformance with Regulatory Guide 1.183 Section C, includes a separate row for Footnote 11 in which the requested exception for the FHA is re-iterated.

The locked rotor dose analysis does not need to consider the high burnup gap fractions used in the FHA. The design of the core ensures that the rods that may be damaged because they may experience departure from nucleate boiling (DNB) following a locked rotor are not those that exceed the linear heat rate and burnup applicability limits of RG 1.183, Revision 0, Table 3, Footnote 11. This is confirmed on a reload-specific basis.

The control rod ejection accident (CREA) dose analysis does not need to consider the high burnup gap fractions used in the FHA. RG 1.183, Revision 0, Table 3, Footnote 11 specifies limitations on linear heat rate and burnup for the Table 3 gap fractions. Footnote 11 and Appendix H, Regulatory Position 1, subsequently identify gap fractions of 10% for iodine and noble gases as applicable to the CREA. The CREA gap fractions are understood to be independent of the Table 3 gap fractions, and thus are not subject to the Footnote 11 limitations on linear heat rate and burnup, which only apply to the Table 3 gap fractions. Thus, the CREA gap fractions do not need to be increased to account for fuel that may exceed the Footnote 11 applicability limits. Note that, in the absence of guidance related to the treatment of alkali metals and other halogens in the CREA dose analysis, the RG 1.183, Revision 0, Table 3 gap fractions of 12% for alkali metals and 5% for other (non-iodine) halogens have been used for the CREA analysis.

Similar approaches with respect to non-LOCA gap fraction selections were taken with Wolf Creek (ML19100A122), Point Beach (ML110240054), Indian Point 2 (ML042960007), Indian Point 3 (ML050750431), and Kewaunee (ML070430020). In these examples, the FHA analysis considered some portion of the fuel exceeding the Footnote 11 limitation while maintaining the Table 3 gap fractions for locked rotor and the subsequent Footnote 11 gap fractions for CREA.

E-16

Vogtle Electric Generating Plant - Units 1 and 2 Response to Request for Additional Information Regarding License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment to Enclosure Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

Attachment to Enclosure Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

RG Section RG Position VEGP Analysis H-1 Assumptions acceptable to the NRC staff Conforms - See discussions in Table regarding core inventory are in Regulatory A. The fission product release is based Position 3 of this guide. For the rod ejection upon Appendix H, the amount of accident, the release from the breached fuel is damaged fuel, and the assumption based on the estimate of the number of fuel rods that 10% of the core inventory of noble breached and the assumption that 10% of the gases and iodine isotopes are in the core inventory of the noble gases and iodines is fuel rod gap. In the absence of specific in the fuel gap. The release attributed to fuel guidance regarding their inclusion, melting is based on the fraction of the fuel that alkali metals (12%) and other reaches or exceeds the initiation temperature for halogens (5%) are conservatively fuel melting and the assumption that 100% of the assumed to be in the fuel gap at levels noble gases and 25% of the iodines contained in consistent with Table 3.

that fraction are available for release from containment. For the secondary system release For releases from containment pathway, 100% of the noble gases and 50% of involving fuel melting, 100% of the the iodines in that fraction are released to the noble gases and 25% of the iodine reactor coolant. isotopes contained in the portion of the fuel that melts is available for release from containment. For releases to the RCS and to the environment through the secondary side, 100% of the noble gases and 50% of the iodines are assumed to be released from melted fuel.

H-2 If no fuel damage is postulated for the limiting Not Applicable - Failed fuel is event, a radiological analysis is not required as postulated for this event.

the consequences of this event are bounded by the consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam generator tube rupture.

A-1

Attachment to Enclosure Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

RG Section RG Position VEGP Analysis H-3 Two release cases are to be considered. In the Conforms - Two release pathways are first, 100% of the activity released from the fuel considered. In the release from should be assumed to be released containment, 100% of the activity from instantaneously and homogeneously through the fuel melting, fuel cladding damage, containment atmosphere. In the second, 100% of and initial RCS inventory the activity released from the fuel should be instantaneously reaches the assumed to be completely dissolved in the containment at the onset of the primary coolant and available for release to the accident and is available for release to secondary system. the environment. In the case with the release from the secondary system, 100% of the activity from fuel melting, fuel cladding damage, and initial RCS inventory instantaneously reaches the RCS at the onset of the accident and is available for release to the secondary system and eventually to the environment.

H-4 The chemical form of radioiodine released to the Conforms - The chemical form of containment atmosphere should be assumed to radioiodine released to the be 95% cesium iodide (CsI), 4.85% elemental containment atmosphere is assumed iodine, and 0.15% organic iodide. If containment to be 95% cesium iodide, 4.85%

sprays do not actuate or are terminated prior to elemental iodine, and 0.15% organic accumulating sump water, or if the containment iodide. Since containment sprays are sump pH is not controlled at values of 7 or not assumed to be activated in this greater, the iodine species should be evaluated event, no credit is taken for pH being on an individual case basis. Evaluations of pH controlled at values of 7 or greater.

should consider the effect of acids created Sump pH is controlled at values >7 during the rod ejection accident event, e.g., with Trisodium phosphate (TSP).

pyrolysis and radiolysis products. With the Therefore, the Position H-4 iodine exception of elemental and organic iodine and species distribution is appropriate.

noble gases, fission products should be assumed to be in particulate form.

H-5 Iodine releases from the steam generators Conforms - The chemical form of iodine to the environment should be assumed to releases from the SGs to the be 97% elemental and 3% organic. environment are assumed to be 97%

elemental and 3% organic.

A-2

Attachment to Enclosure Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

RG Section RG Position VEGP Analysis H-6.1 A reduction in the amount of radioactive material Conforms - Radioactive material available for leakage from the containment that is removal from the containment due to natural deposition, containment sprays, atmosphere by sprays and other recirculating filter systems, dual containments, or engineered safety features is not other engineered safety features may be taken credited. Plateout of elemental iodine is into account. Refer to Appendix A to this guide for not credited. Natural deposition of guidance on acceptable methods and aerosols is credited by using the assumptions for evaluating these mechanisms. bounding removal coefficient calculated from NUREG/CR-6189 Table 36.

NUREG/CR-6189 is an acceptable model for aerosol natural deposition per RG 1.183 Appendix A.

H-6.2 The containment should be assumed to leak at Conforms - The containment is the leak rate incorporated in the technical assumed to leak to the environment at specifications at peak accident pressure for the the technical specification limit plus first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the 5% (0.21%/day) for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> remaining duration of the accident. Peak of the accident and half this rate accident pressure is the maximum pressure thereafter.

defined in the technical specifications for containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.

H-7.1 A leak rate equivalent to the primary-to- Conforms - The total leakage from the secondary leak rate limiting condition for primary system to the secondary operation specified in the technical specifications system is assumed to be 1 gpm, should be assumed to exist until shutdown conservatively bounding the technical cooling is in operation and releases from the specification limit of 150 gpd per steam generators have been terminated. generator. This leakage lasts for the entire duration of steam release from the secondary side (20 hrs).

H-7.2 The density used in converting volumetric leak Conforms - The water density of both rates (e.g., gpm) to mass leak rates (e.g., the primary and secondary coolants is lbm/hr) should be consistent with the basis of assumed to be 62.4 lbm/ft3.

surveillance tests used to show compliance with leak rate technical specifications. These tests typically are based on cooled liquid. The facilitys instrumentation used to determine leakage typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

A-3

Attachment to Enclosure Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident)

RG Section RG Position VEGP Analysis H-7.3 All noble gas radionuclides released to the Conforms - It is assumed that noble secondary system are assumed to be released gases are not retained in the to the environment without reduction or secondary water.

mitigation.

H-7.4 The transport model described in Conforms - The transport model assumptions 5.5 and 5.6 of Appendix E described in Position 5.5 and 5.6 of should be utilized for iodine and Appendix E is applied to releases from particulates. the steam generators.

A-4