Text

License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490
	ML22181B066
Person / Time
Site:	Vogtle
Issue date:	06/30/2022
From:	Gayheart C; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-22-0208
	Download: ML22181B066 (223)
	v • d • e

Cheryl A. Gayheart 3535 Colonnade Parkway Regulatory Affairs Director Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@southernco.com June 30, 2022 Docket Nos.: 50-424 NL-22-0208 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Ladies and Gentlemen:

Pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations, Southern Nuclear Operating Company (SNC) hereby submits license amendment requests (LARs) to renewed facility operating licenses NPF-68 and NPF-81 to revise the Technical Specifications (TS) for the Vogtle Electric Generating Plant (VEGP), Units 1 and 2, respectively.

SNC requests Nuclear Regulatory Commission (NRC) review and approval of the proposed revisions to the licensing basis of VEGP that support a selected scope application of an Alternative Source Term (AST) methodology. The proposed changes, which are supported by the AST Design Basis Accident radiological consequences analysis, are included in this license amendment request (LAR). In addition, the proposed amendment incorporates 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 Tech Spec, Revision 0.

The Enclosure to this letter contains SNCs evaluation of the proposed changes. Attachments to this enclosure provide supporting documentation, as required. Attachment 1 of the Enclosure provides the existing TS pages marked-up to show the proposed changes. Note that existing TS pages 3.4.16-3 and 3.4.16-4 are proposed to be deleted in their entirety.

SNC requests approval of this LAR one year from acceptance. The proposed changes would be implemented within 120 days of issuance of the amendment.

This document contains NRC regulatory commitments as stated in Attachment 12 of the Enclosure.

In accordance with 10 CFR 50.91(b)(1), SNC is notifying the State of Georgia of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.

If you have any questions regarding this submittal, please contact Ryan Joyce at 205.992.6468.

Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Enclosure Basis for Proposed Change

Enclosure to NL-22-0208 Basis for Proposed Change Enclosure Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description

2.1 Background

2.2 TSTF-51 2.3 TSTF-471 2.4 TSTF-490 3.0 Technical Evaluation 3.1 Meteorology and Atmospheric Dispersion 3.2 Radionuclide Inventory 3.3 Analytical Models 3.4 Loss of Coolant Accident 3.5 Fuel Handling Accident 3.6 Main Steam Line Break Accident 3.7 Steam Generator Tube Rupture Accident 3.8 Control Rod Ejection Accident 3.9 Locked Rotor Accident 3.10 Equipment Qualification 3.11 Accident Analysis 3.12 Summary of FHA Dose Results 3.13 Acceptability of the Proposed Change 3.14 Variations from TSTF-51 and TSTF-471 3.15 TSTF-490 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions E-1

Enclosure to NL-22-0208 Basis for Proposed Change Enclosure Basis for Proposed Change Table of Contents (continued) 5.0 Environmental Consideration 6.0 References Attachments A1 Technical Specification Pages (Markup)

A2 Technical Specification Pages (Clean Typed Pages)

A3 Technical Specification Bases Pages (Markup) (For information only)

A4 Regulatory Guide 1.183 Conformance Tables A5 Loss-of-Coolant Accident Analysis A6 Fuel Handling Accident Analysis A7 Main Steam Line Break Accident Analysis A8 Steam Generator Tube Rupture Accident Analysis A9 Control Rod Ejection Accident Analysis A10 Locked Rotor Accident Analysis A11 VEGP AST Accident Analysis Input Values Comparison Tables A12 List of Regulatory Commitments E-2

Enclosure to NL-22-0208 Basis for Proposed Change 1.0 Summary Description This evaluation supports a request to revise Operating Licenses NPF-68 and NFP-81 for Vogtle Electric Generating Plant (VEGP), Units 1 and 2, respectively. 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 selective scope application of an Alternative Source Term (AST) methodology. The proposed amendment also incorporates Revision 2 of Technical Specification Task Force (TSTF)

Traveler, TSTF-51, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations, Revision 1 of TSTF-471, Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes, and Revision 0 of TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec, into the VEGP Technical Specifications.

2.0 Detailed Description

2.1 Background

In December 1999, the NRC issued a new regulation, 10 CFR 50.67, "Accident Source Term," which provided a mechanism for licensed power reactors to voluntarily replace the traditional accident source term used in their Design Basis Accident (DBA) analyses with an AST. Regulatory guidance for the implementation of the AST is provided in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Reference 1). 10 CFR 50.67 requires a licensee seeking to use an AST to submit a license amendment request (LAR) and requires that the application contain an evaluation of the consequences of DBAs.

This LAR addresses the applicable requirements and guidance in proposing to use an AST in evaluating the offsite and Control Room (CR) radiological consequences of the VEGP DBAs. This reanalysis involves several changes in selected analysis assumptions. As part of the implementation of the AST, the Total Effective Dose Equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

This will also replace the whole body (and its equivalent to any part of the body) dose criteria of 10 CFR 50, Appendix A, General Design Criterion (GDC) 19.

2.2 TSTF-51 The proposed amendment would revise certain Technical Specifications (TSs) to remove the requirements for certain engineered safety features (ESF) systems to operate after sufficient radioactive decay of irradiated fuel has occurred following a plant shutdown. Following sufficient radioactive decay, these ESF systems are no longer required during a fuel handling accident (FHA) to ensure main control E-3

Enclosure to NL-22-0208 Basis for Proposed Change room (MCR) personnel dose remains below the 10 CFR 50.67(b)(2)(iii) dose limit and of-site dose remains below the accident dose limit specified in the NRC standard review plan, which represents a small fraction of the 10 CFR 50.67 does limits.

Associated with this change is the deletion of OPERABILITY requirements during CORE ALTERATIONS for certain ESF mitigation features. This change will allow flexibility to move personnel and equipment and perform work which would affect containment OPERABILITY during the handling of irradiated fuel.

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed changes are based on performing analyses assuming a longer decay period to take advantage of the reduced radionuclide inventory available for release in the event of an FHA. Following sufficient decay occurring, the primary success path for mitigating the FHA no longer includes the functioning of the active containment systems. Therefore, the OPERABILITY requirements of the TS are modified to reflect that water level and decay time are the primary success path for mitigating an FHA (which meets General Design Criterion 3).

To support this change in requirements during the handling of irradiated fuel, the OPERABILITY requirements during CORE ALTERATIONS for certain ESF mitigation features are deleted. The accidents postulated to occur during core alterations, in addition to fuel handling accidents, are: inadvertent criticality (due to control rod removal error or continuous control rod withdrawal error during refueling or boron dilution) and the inadvertent loading of, and subsequent operation with, a fuel assembly in an improper location. These events are not postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur during CORE ALTERATIONS that results in a significant radioactive release is the FHA, the proposed TS requirements omitting CORE ALTERATIONS is justified.

Also, the TS only allow the handling of irradiated fuel in the reactor vessel when the water level in the reactor cavity is at the high water level. Therefore, the proposed changes only affect containment requirements during periods of relatively low shutdown risk during refueling outages. Therefore, the proposed changes do not significantly increase the shutdown risk.

2.3 TSTF-471 Suspending CORE ALTERATIONS or exempting testing except during CORE ALTERATIONS has no effect on the initial conditions or mitigation of any Design Basis Accident (DBA) or transient, and these requirements apply an operational burden with no corresponding safety benefit. Therefore, the proposed amendment eliminates the use of the defined term CORE ALTERATIONS from only the following select Technical Specifications.

TS 3.9.1 Boron Concentration E-4

Enclosure to NL-22-0208 Basis for Proposed Change TS 3.9.2 Unborated Water Source Isolation Valves TS 3.9.4 Containment Penetrations 2.4 TSTF-490 The proposed amendment would replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-133 and would take into account only the noble gas activity in the primary coolant.

The background for this proposed change is as stated in the model SE in the NRC Notice of Availability published on March 19, 2007 (72 FR 12838)

(Reference 2), the NRC Notice for Comment published on November 20, 2006 (71 FR 67170 (Reference 3), and TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.

3.0 Technical Evaluation Alternative Source Term 3.1 Meteorology and Atmospheric Dispersion The AST application uses atmospheric dispersion (/Q) values for the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ), and the CR receptors.

As described below, the EAB and LPZ /Q values are consistent with the current licensing basis, as given in VEGP Final Safety Analysis Report (FSAR)

Table 2.3.4-1 and Table 15A-2. New and revised /Q values for the CR have been developed to address potential leakage from the Refueling Water Storage Tank (RWST) vent and releases from the secondary side for evaluation of non-LOCA radiological consequences. The resulting /Q values at the CR intakes are calculated using the NRC-sponsored computer code ARCON96 (NUREG/CR-6331) consistent with the procedures in RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," (Reference 4). Information used to develop the new

/Q values is included in Attachment 11.

3.1.1 Meteorological Data The same meteorological data used to calculate the /Q values applied in the current licensing basis radiological consequence analyses was determined to remain representative of the site and was used to calculate new CR /Q values.

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Enclosure to NL-22-0208 Basis for Proposed Change 3.1.2 EAB and LPZ Atmospheric Dispersion Factors RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," Section 5.3, "Meteorology Assumptions," states:

Atmospheric dispersion values (/Q) for the EAB, the LPZ, and the CR that were approved by the staff during the initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide.

For the AST analyses, /Q values for the EAB and the LPZ are consistent with the current licensing basis.

The /Q values for the EAB and the LPZ used in the radiological consequence analyses are shown in Table 3.1.

Table 3.1 - EAB and LPZ /Q values (sec/m3)

Location Time Period /Q Value EAB 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.8x10-4 LPZ 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 7.2x10-5 LPZ 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.3x10-5 LPZ 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.2x10-5 LPZ 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 9.2x10-6 LPZ 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.7x10-6 3.1.3 Control Room Atmospheric Dispersion Factors

/Q values for onsite release-receptor combinations were developed using the ARCON96 computer code and the guidance in RG 1.194.

Various release-receptor combinations were considered for the onsite CR

/Q values. These different cases are considered to determine the limiting release-receptor combination for the events postulated in RG 1.183. Existing /Q values included releases from the Containment, Containment Hatch Door, and Fuel Handling Building. New /Q values were developed for the RWST release points for the Loss-of-Coolant Accident (LOCA) as well as the North and South Main Steam Valve Rooms for secondary side releases in non-LOCA events such as the Control Rod Ejection Accident (CREA), Locked Rotor Accident (LRA),

Steam Generator Tube Rupture (SGTR), and Main Steam Line Break (MSLB).

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Enclosure to NL-22-0208 Basis for Proposed Change Figure 3.1 provides a sketch of the general layout of Vogtle 1 & 2 that has been annotated to highlight the onsite release and receptor point locations. All releases are taken as ground level releases per RG 1.194 Position 3.2.1.

Table 3.2 provides information related to the relative elevations of the release-receptor combinations, the straight-line horizontal distance between the release point and the receptor location, and the direction (azimuth) from the receptor location to the release points. Angles are calculated based on trigonometric layout of release and receptor points in relation to the North-South and East-West axes.

Figure 3.1 - Air Intake Locations and Release Points E-7

Enclosure to NL-22-0208 Basis for Proposed Change Table 3.2 - Distance and Geometry of Release and Receptor Locations Receptor: Unit 1 CR Air Intake Modeled Release Direction Horizontal Horizontal Modeled Release Release Point Elevation above to Source Distance (ft) Distance (m) Elevation (ft)

Grade (degree)

Unit 1 (ft) (m)

Containment 70.9 21.6 283.0* 63.0 19.2 136 RWST 365.0 111.2 283.0* 63.0 19.2 160 North Release 91.9 28.0 283.0* 63.0 19.2 90 South Release 207.6 63.3 283.0* 63.0 19.2 151 Plant Vent 144.6 44.1 420.1 200.1 61.0 136 Hatch Door 275.1 83.8 228.5 8.5 2.6 135 FHB 106.0 32.3 288.2 68.2 20.8 180 Unit 2 Containment 188.1 57.3 283.0* 63.0 19.2 246 RWST 432.4 131.8 283.0* 63.0 19.2 217 North Release 232.0 70.7 283.0* 63.0 19.2 270 South Release 301.2 91.8 283.0* 63.0 19.2 233 Plant Vent 261.8 79.8 420.1 200.1 61.0 246 Hatch Door 363.0 110.6 228.5 8.5 2.6 243 Receptor: Unit 2 CR Air Intake Horizontal Modeled Release Direction Horizontal Modeled Release Release Point Distance Elevation above to Source Distance (m) Elevation (ft)

(ft) Grade (degree)

Unit 1 (ft) (m)

Containment 188.1 57.3 283.0* 63.0 19.2 114 RWST 432.4 131.8 283.0* 63.0 19.2 143 North Release 232.0 70.7 283.0* 63.0 19.2 90 South Release 301.2 91.8 283.0* 63.0 19.2 127 Plant Vent 261.8 79.8 420.1 200.1 61.0 114 Hatch Door 363.0 110.6 228.5 8.5 2.6 118 Unit 2 Containment 70.9 21.6 283.0* 63.0 19.2 224 RWST 365.0 111.2 283.0* 63.0 19.2 200 North Release 91.9 28.0 283.0* 63.0 19.2 270 South Release 207.6 63.3 283.0* 63.0 19.2 209 Plant Vent 144.6 44.1 420.1 200.1 61.0 224 Hatch Door 275.1 83.8 228.5 8.5 2.6 225 FHB 106.0 32.3 288.2 68.2 20.8 180

Release heights are conservatively made the same as the Control Room intakes.

Table 3.3 provides the ARCON96 modeling outputs for releases originating at the Containments, reactor vents, hatch doors, RWSTs, FHB, and North and South Main Steam Valve Rooms. The individual AST analyses use /Q values that bound both units and correspond to the limiting release points applicable to the event. Refer to the individual accident Attachments for the CR /Q values used in each analysis.

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Enclosure to NL-22-0208 Basis for Proposed Change Table 3.3 - /Q (s/m3) Values at the Control Room Air Intakes Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 - 4 days 4 - 30 days U1 Reactor U1 CR 2.20E-03 1.31E-03 5.26E-04 4.64E-04 3.32E-04 U1 Hatch Door U1 CR 6.82E-04 4.85E-04 1.75E-04 1.43E-04 1.03E-04 U1 RWST U1 CR 5.23E-04 3.81E-04 1.44E-04 1.15E-04 8.48E-05 U1 North U1 CR 7.64E-03 6.17E-03 2.57E-03 1.74E-03 1.33E-03 U1 South U1 CR 1.57E-03 1.12E-03 4.10E-04 3.32E-04 2.41E-04 U1 Vent U1 CR 1.70E-03 1.23E-03 4.30E-04 3.50E-04 2.43E-04 U2 Reactor U1 CR 8.46E-04 6.63E-04 2.96E-04 2.36E-04 1.71E-04 U2 Hatch Door U1 CR 4.42E-04 3.64E-04 1.61E-04 1.28E-04 9.07E-05 U2 RWST U1 CR 4.02E-04 3.21E-04 1.30E-04 1.08E-04 7.88E-05 U2 North U1 CR 1.31E-03 1.05E-03 4.70E-04 3.22E-04 2.63E-04 U2 South U1 CR 8.31E-04 6.62E-04 2.76E-04 2.31E-04 1.57E-04 U2 Vent U1 CR 7.71E-04 5.59E-04 2.35E-04 1.96E-04 1.41E-04 FHB (east) U1 CR 6.01E-03 4.44E-03 1.71E-03 1.40E-03 1.07E-03 U1 Reactor U2 CR 8.01E-04 5.52E-04 2.34E-04 1.78E-04 1.43E-04 U1 Hatch Door U2 CR 4.30E-04 3.06E-04 1.19E-04 9.11E-05 6.85E-05 U1 RWST U2 CR 3.74E-04 2.65E-04 9.49E-05 7.91E-05 5.71E-05 U1 North U2 CR 1.31E-03 1.06E-03 4.43E-04 3.02E-04 2.32E-04 U1 South U2 CR 7.83E-04 5.30E-04 1.98E-04 1.52E-04 1.12E-04 U1 Vent U2 CR 7.91E-04 5.59E-04 2.16E-04 1.60E-04 1.17E-04 U2 Reactor U2 CR 2.22E-03 1.55E-03 6.57E-04 5.80E-04 4.47E-04 U2 Hatch Door U2 CR 7.16E-04 5.83E-04 2.45E-04 2.05E-04 1.42E-04 U2 RWST U2 CR 5.42E-04 4.27E-04 1.69E-04 1.37E-04 1.01E-04 U2 North U2 CR 7.58E-03 6.10E-03 2.72E-03 1.86E-03 1.52E-03 U2 South U2 CR 1.67E-03 1.32E-03 5.18E-04 4.21E-04 3.15E-04 U2 Vent U2 CR 1.67E-03 1.20E-03 5.16E-04 4.39E-04 3.25E-04 FHB (west) U2 CR 6.01E-03 4.44E-03 1.71E-03 1.40E-03 1.07E-03 Notes: U1 - Unit 1; U2 - Unit 2; CR - Control Room; FHB - Fuel Handling Building; RWST - Refueling Water Storage Tank 3.2 Radionuclide Inventory 3.2.1 Fission Product Inventory 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 GWD/MTU and variable enrichment regions. Parametric studies were performed varying enrichment (up to 5 weight percent), lead rod burnup, and beginning and end of cycle burnup. The fission product inventory used in the AST analyses applies a fuel design margin to the nominal inventory that ensures the AST dose consequences bound the dose consequences of the parametric study maxima.

3.2.2 Equilibrium RCS and Secondary Source Terms For the AST dose analyses that consider it, the equilibrium RCS source term is assumed to consist of halogens (I, Br), noble gases (Kr, Xe), and alkali metals (Cs, Rb). The concentrations of I-131, I-132, I-133, I-134, and I-135 are assumed to be based on 1.0 µCi/g dose equivalent (DE)

I-131, consistent with the Technical Specification 3.4.16 limit and revised E-9

Enclosure to NL-22-0208 Basis for Proposed Change definition of DE I-131 from TSTF-490. The concentrations of Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 are be based on 280 µCi/g DE Xe-133, consistent with the proposed Technical Specification 3.4.16B limit and definition of DE Xe-133 from TSTF-490. Remaining nuclides are assumed to be present in the equilibrium RCS source term at concentrations corresponding to 1% of the fuel containing cladding defects.

For the AST dose analyses that consider it, the equilibrium secondary side source term is assumed to consist of halogens (I, Br) and alkali metals (Cs, Rb). The concentrations of I-131, I-132, I-133, I-134, and I-135 are based on 0.1 µCi/g dose equivalent (DE) I-131, consistent with Technical Specification LCO 3.7.16. The concentrations of other halogens and alkali metals in the equilibrium secondary side source term are assumed to be 10% of the RCS equilibrium concentrations, based on the ratio of the secondary to RCS DE I-131 concentrations.

The individual accident sections and Attachments describe the RCS and secondary side source term considerations applicable to that event (e.g.,

iodine spiking in the SGTR and MSLB).

3.3 Analytical Models The following computer codes are used in performing the VEGP AST radiological dose analyses:

RADTRAD is used to determine the CR and offsite doses for the LOCA, FHA, CREA, LRA, SGTR, and MSLB using the source term, /Q inputs, and pertinent scenario specific requirements in Regulatory Guide 1.183 as described in the individual event sections and Attachments. The code considers the radionuclide release timing and chemical form, hold-up and removal (e.g., spray, deposition) within compartments, filtration, transport to receptors, and dose calculation.

ARCON96 was used to determine the /Qs at the CR intakes for selected release locations using plant meteorological data.

ORIGEN-ARP was used to calculate bounding plant-specific fission product inventories for use in the dose calculations that postulate fuel damage (LOCA, FHA, CREA, LRA).

3.4 Loss of Coolant Accident The LOCA is a postulated rupture in the reactor coolant system that results in expulsion of the coolant to containment. Even though the emergency core cooling system (ECCS) is designed to maintain cooling of the fuel assemblies in this event, the dose consequence analysis is performed assuming a significant release of the radionuclides from the fuel assemblies.

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Enclosure to NL-22-0208 Basis for Proposed Change 3.4.1 Methodology Overview The LOCA is modeled as a release of nuclides from the reactor core into the containment building. The Containment release paths modeled are:

1) the Containment Mini-Purge System,

2) Containment leakage,

3) ECCS leakage, and

4) RWST backleakage.

The radiological source term characteristics and release timing are based on the AST methodology in RG 1.183. Atmospheric dispersion factors from Section 3.1, above, are used in this analysis. Doses to the public at the EAB and the LPZ, and occupants in the CR are determined.

3.4.2 Radiological Dose Models The RADTRAD (Version 3.10) code was used to calculate the immersion and inhalation dose contributions to both the onsite and offsite radiological dose consequences. Models were developed for both the containment leakage and ECCS leakage cases.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Attachment 5. The calculated dose results are given in Table 3.4. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a LOCA. These TEDE criteria are 25 rem at the EAB (worst 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months interval) and LPZ (cumulative for the 30 day accident duration),

and 5 rem in the CR (cumulative for the 30 day accident duration).

Table 3.4 - Calculated LOCA Radiological Consequences TEDE (rem)

EAB* LPZ Control Room Calculated results 8.4 9.6 4.4 Dose acceptance criteria 25 25 5

Worst 2-hour dose 3.5 Fuel Handling Accident The limiting case of a fuel handling accident (FHA) in the fuel building is analyzed.

Analysis of the FHA in the fuel building takes no credit for either filtration, or holdup in the fuel building. The dispersion factors for a release from the fuel building are much larger than those for a release from the containment E-11

Enclosure to NL-22-0208 Basis for Proposed Change building (containment open configuration) equipment hatch and personnel air locks (PALs). In addition, a release through the containment PAL would result in a torturous path to the CR through the auxiliary building out to the atmosphere and finally into the CR from the control room emergency filtration system (CREFS) supply intake.

For the containment closed case where the purge system is operating, the purge flow rates would result in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months release duration as compared to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months duration required by RG 1.183 Revision 0 for a release from the fuel building. This release pathway would discharge through the plant vent which has a much smaller dispersion factor as compared to that of the fuel handling building. For these reasons, the FHA in the fuel building is bounding compared to any release from containment (open or closed). Certain sections in the VEGP FSAR are retained for historical purposes, but it is important to note that an FHA inside containment is not limiting regardless of containment configuration compared to a release from the fuel building.

Table 3.5a shows a comparison of atmospheric dispersion factors which illustrates this further.

High radiation in the CR makeup air intake results in isolation of the CR. A delay time of 10 minutes is conservatively assumed for radiation levels to reach the CREFS actuation setpoint.

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. This is consistent with the LHGR limit depicted in Figure A.1 of PNNL-18212 Revision 1. To demonstrate that this request is acceptable, the gap release fractions assumed in the FHA analysis were taken from Table 2.9 of PNNL-18212 Revision 1. These values reflect increases in release fractions for Kr-85, I-132, and other Noble Gases. For conservatism, the source term developed for the FHA analysis assumed 100% of the failed rods exceeded the RG 1.183 footnote 11 limit. The source term at 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months after reactor shutdown used in the FHA analysis is provided in Attachment 6.

The calculated dose results are given in Table 3.5b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for an FHA.

These TEDE criteria are 6.3 rem at the EAB for the worst two hours, 6.3 rem at the LPZ for the duration of the accident (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) and 5 rem in the CR for the duration of the accident.

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Enclosure to NL-22-0208 Basis for Proposed Change Table 3.5a: Comparison of X/Qs (s/m3) for the FHA Analysis Release Receptor 0-2 Hours 2-8 Hours 8-24 Hours U1 Hatch Door U1 MCR 6.82E-04 4.85E-04 1.75E-04 U1 Vent U1 MCR 1.70E-03 1.23E-03 4.30E-04 FHB (east) U1 MCR 6.01E-03 4.44E-03 1.71E-03 U2 Hatch Door U1 MCR 4.42E-04 3.64E-04 1.61E-04 U2 Vent U1 MCR 7.71E-04 5.59E-04 2.35E-04 U1 Hatch Door U2 MCR 4.30E-04 3.06E-04 1.19E-04 U1 Vent U2 MCR 7.91E-04 5.59E-04 2.16E-04 FHB (west) U2 MCR 6.01E-03 4.44E-03 1.71E-03 U2 Hatch Door U2 MCR 7.16E-04 5.83E-04 2.45E-04 U2 Vent U2 MCR 1.67E-03 1.20E-03 5.16E-04 Max 6.01E-03 4.44E-03 1.71E-03 Table 3.5b: FHA Analysis Results Location/Dose Point TEDE (REM) Acceptance Criteria (REM)

Exclusion Area Boundary (EAB) 1.0 6.3 Low Population Zone (LPZ) 0.4 6.3 Control Room (CR) 3.9 5 E-13

Enclosure to NL-22-0208 Basis for Proposed Change 3.6 Main Steam Line Break Accident This event consists of a break in one main steam line outside of containment in which the faulted steam generator (SG) completely depressurizes and instantly releases the initial contents of the faulted SG secondary side to the environment.

The plant cooldown continues by dumping steam with the intact SGs. In addition to the release of nuclides that are initially present in the SG secondary side, leakage of primary coolant into the SG secondary side occurs at a rate equal to 0.35 gpm to the faulted SG, and 0.65 gpm to the intact SGs (1.0 gpm total). This is conservative relative to the TS limit of 150 gallons per day per SG.

Two iodine spike cases are considered. In the pre-accident iodine spike case, a reactor transient is assumed to occur prior to the event in which the primary coolant iodine concentration has increased to the maximum TS value of 60 Ci/gm. For the concurrent spike case, the initial primary iodine activity release rate is at 500 times the equilibrium TS value of 1.0 Ci/g Dose Equivalent Iodine. This concurrent iodine spike is assumed to have a duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . No fuel damage is postulated to occur for the MSLB event.

Leakage from the RCS into all of the SGs, and steam release from the intact SGs, continues until the RCS is placed on RHR cooling after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . The leakage to the faulted SG is modeled as a direct flow from the RCS to the environment without partitioning. In the leakage to the intact SGs, noble gases are assumed to leak directly to the environment. A partition factor of 100 is applied to the iodine nuclides in the intact SGs.

The release locations from the faulted and intact SGs are conservatively taken as the most limiting MSIV area release locations. The CR is automatically realigned into the emergency ventilation mode upon receipt of a safety injection signal.

The analysis used assumptions and inputs that follow the guidance in RG 1.183.

The key parameters and assumptions are listed in Attachment 7. The calculated dose results are given in Table 3.6. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a MSLB. These TEDE criteria are 25 rem at the EAB and LPZ for the fuel damage or pre-incident spike case, and 2.5 rem at the EAB and LPZ for the concurrent iodine spike case. The TEDE criteria is 5 rem for the CR occupant in both cases. The duration of the release is until the residual heat removal (RHR) system is placed in service (20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ); the resultant doses are calculated for 30 days post-accident.

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Enclosure to NL-22-0208 Basis for Proposed Change Table 3.6 - Calculated MSLB Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results Pre-Incident Spike <0.1 <0.1 <0.1 Concurrent Iodine Spike 0.2 0.2 0.2 TEDE (rem)

EAB LPZ Control Room Dose acceptance criteria Fuel Damage or Pre-Incident Spike 25 25 5 Concurrent Iodine Spike 2.5 2.5 5 3.7 Steam Generator Tube Rupture Accident The SGTR event represents an instantaneous rupture of a SG tube that releases primary coolant into the lower pressure secondary system. In addition to the break flow rate, primary-to-secondary leakage (consistent with CLB) occurs at a rate equal to 0.3 gpm to the ruptured and 0.7 gpm for the intact SGs (1.0 gpm total). Leakage into the SGs continues until the RCS is placed on residual heat removal (RHR) system cooling at 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

A portion of the break flow to the ruptured SG flashes to vapor based upon the thermodynamic conditions in the reactor and secondary coolant. The portion of the primary coolant that does flash in the SG secondary is released directly to the environment without mitigation. The break flow that does not flash mixes with the bulk water in the SG where the activity is released based upon the steaming rate and a partition factor. A SG partition factor of 100 is applied to the iodine nuclides.

Two iodine spike cases are considered. In the pre-accident iodine spike case, a reactor transient is assumed to occur prior to the event in which the primary coolant iodine concentration has increased to the maximum TS value of 60 Ci/gm. For the concurrent spike case, the initial primary iodine activity concentration is at the equilibrium TS value of 1.0 Ci/g Dose Equivalent Iodine.

This concurrent iodine spike is assumed to have a duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In both cases, as an initial condition, RCS activity includes an equilibrium noble gas concentration of 280 µCi/g dose equivalent Xe-133 and consideration of clad defects in 1% of the fuel rods for the other radioactive nuclides.

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Enclosure to NL-22-0208 Basis for Proposed Change The release locations from the faulted and intact SGs are conservatively taken as a release from the main steam line room closest to the CR. This is conservative as this is where the main steam safety relief valves and (power operated) atmospheric relief valves (ARVs) are located. The CR is automatically realigned into the emergency ventilation mode upon receipt of a safety injection signal.

The SG blowdown sample lines are assumed open in the analysis and discharge from the faulted and intact steam generators for the duration of the event. There is no credit for partitioning of nobles, iodines, or particulates from the blowdown sample line flow. All of the blowdown sample line flow is assumed to flash and be released from the limiting location in the main steam line room in the auxiliary building (discussed above) for conservatism.

The analysis used assumptions and inputs that follow the guidance in RG 1.183.

The key parameters and assumptions are listed in Attachment 8. The calculated dose results are given in Table 3.7. The analysis used a release duration of 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months and an exposure duration of 30 days. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a SGTR. These TEDE criteria are 25 rem at the EAB and LPZ for the pre-accident spike case, and 2.5 rem at the EAB and LPZ for the concurrent iodine spike case. The TEDE criteria is 5 rem for the CR occupant in both cases.

Table 3.7 - Calculated SGTR Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results Pre-Accident Spike 1.6 0.9 0.6 Concurrent Spike 1.4 0.8 0.5 TEDE (rem)

EAB LPZ Control Room Dose acceptance criteria Pre-Accident Spike 25 25 5 Concurrent Spike 2.5 2.5 5 E-16

Enclosure to NL-22-0208 Basis for Proposed Change 3.8 Control Rod Ejection Accident The Control Rod Ejection event involves a reactivity insertion that produces a short, rapid core power level increase, which results in fuel rod damage and localized melting. Two separate release pathways are evaluated independently:

a release from containment and a release from the secondary system. In both cases, 10% of the noble gases and 10% of the iodine isotopes in the core are available for release from the fuel gap of the damaged fuel rods. In addition, 12%

of the alkali metals and 5% of the other halogens (i.e., Br) are also assumed to be located in the fuel rod gap. It is assumed that the reactor coolant contains the maximum equilibrium iodine concentration (1.0 µCi/gm DE I-131) as well as noble gases based on 280 µCi/gm DE Xe-133 and other radionuclides corresponding to 1% defective fuel. The initial secondary side activity is assumed to be 10% of the primary side activity.

For releases from containment, 10% of the fuel rods in the core experience cladding failure and 0.25% of the fuel experiences melting. The activity in the fuel rod gap of the damaged fuel is instantaneously and uniformly mixed throughout the containment atmosphere. Moreover, 100% of the noble gases and 25% of the iodine isotopes in the melted fuel are also added to the fission product inventory in containment.

No credit is taken for removal by containment sprays or for deposition of elemental iodine on containment surfaces. Credit is taken for natural deposition of aerosols in containment based on a conservatively low removal coefficient.

Activity is released from containment at the TS leak rate limit plus 5% for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at half that rate after that.

For releases from the secondary system, 10% of the fuel rods in the core are breached and 0.25% of the fuel experiences melting. Activity released from the fuel is completely dissolved in the primary coolant and is available for release to the secondary system. In this case, 100% of the noble gases and 50% of the iodine isotopes in the melted fuel are also released into the reactor coolant. The noble gases are assumed to be released directly to the environment, and the remaining fission products are transported from the RCS to the SGs at 1 gpm, which is conservative relative to the TS limit of 150 gallons per day per SG. The leakage duration is assumed to continue until the termination of steaming from the SGs. The secondary side mass releases are conservatively assumed to last for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

The atmospheric dispersion factors are conservatively taken as the most limiting release and receptor locations for each pathway and each averaging period. The CR ventilation system is automatically realigned into the emergency ventilation mode following receipt of a safety injection signal.

The analysis used assumptions and inputs that follow the guidance in RG 1.183.

The key parameters and assumptions are listed in Attachment 9. The calculated dose results are given in Table 3.8. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a Control Rod Ejection. These E-17

Enclosure to NL-22-0208 Basis for Proposed Change TEDE criteria are 6.3 rem at the EAB and LPZ, and 5 rem for the CR occupant.

The accident release duration is 30 days for the Containment pathway, and until the residual heat removal (RHR) system is placed in service for the secondary pathway.

Table 3.8 - Calculated Control Rod Ejection Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Containment Release 1.5 1.8 0.6 Secondary Release 0.5 0.4 1.1 Dose acceptance criteria 6.3 6.3 5 3.9 Locked Rotor Accident The Locked Rotor Accident dose analysis is defined by the 5% of the fuel rods which become damaged by the event. A radial peaking factor of 1.7 is assumed.

Radionuclides released from the fuel are instantaneously and uniformly mixed throughout the primary coolant. Noble gases are released directly to the environment, and the remaining isotopes are transported to the SGs at a rate of 1 gpm. This continues for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , by which time the RCS is placed on residual heat removal (RHR) system cooling.

The quantity of the fission products released from the failed fuel dominates the RCS activity during the event; however, the initial nuclide concentration in the RCS prior to the event is considered. The analysis also includes the dose contribution from the release of iodine initially present in the SG secondary side.

The release locations are conservatively taken as the most limiting release locations from the MSIV area. The analysis assumes that the CR isolates (608 seconds from transient initiation) on high radiation from the control room intake radiation monitors and then enters the emergency ventilation mode (698 seconds from transient initiation).

The analysis used assumptions and inputs that follow the guidance in RG 1.183.

The key parameters and assumptions are listed in Attachment 10. The calculated dose results are given in Table 3.9. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a Locked Rotor Accident.

These TEDE criteria are 2.5 rem at the EAB and LPZ, and 5 rem for the CR occupant. The duration of the release is 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months and the dose results are calculated for 30 days.

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Enclosure to NL-22-0208 Basis for Proposed Change Table 3.9 - Calculated Locked Rotor Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results <0.1 <0.1 0.3 Dose acceptance criteria 2.5 2.5 5 3.10 Equipment Qualification Impacts to environmental qualification of equipment were evaluated against the legacy approved approach (TID-14844) for Vogtle. The evaluation concluded that environmental qualification of equipment at Vogtle is unaffected by the adoption of RG-1.183.

TSTF- 51 and TSTF-471 3.11 Accident Analysis The control of movement of loads heavier than a fuel assembly over irradiated fuel is described in VEGP FSAR Section 9.1.5. Section 9.1.5 also demonstrates conformance to NUREG-0612 and includes the following topics:

Safe Load Paths Establishment of load handling procedures Implementation of standards with respect to: training of crane operators, use of special lifting devices, use of slings, and design, inspection, testing, and maintenance of cranes Heavy load drop analyses As summarized in FSAR sub-subsection 9.1.5.3.1, "Postulated Loads Inside Containment," and Section 9.1.5.3.2 Postulated Loads Inside Fuel Handling Building the effects of heavy load drops have been evaluated and as presented in FSAR Table 9.1.5-3 satisfy NUREG-0612 criteria. It has been verified that the buckling load on affected fuel assemblies would not exceed design limits and that there will be no consequential damage to the structural integrity of the reactor vessel, reactor vessel nozzles, or RCS loop piping. Therefore, core cooling capability and the integrity of the fuel cladding will be maintained. Thus, the inadvertent drop of a heavy load at VEGP would have no impact on the health or safety of the public. The proposed license amendment does not impact or alter the VEGP load drop analysis described in Subsection 9.1.5 of the FSAR.

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Enclosure to NL-22-0208 Basis for Proposed Change 3.12 Summary of FHA Dose Results A summary of the AST FHA dose results is located Section 3.5 of this Enclosure.

3.13 Acceptability of the Proposed Change Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed change is based on the results of the VEGP FHA analyses that assumes a fuel decay period of at least 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months reducing the radionuclide inventory available for release to the environment in the event of an FHA. Following sufficient decay occurring, the primary success path for mitigating the FHA no longer includes the functioning of the active containment systems to ensure off-site and MCR doses remain below the 10 CFR 50.67 dose limits.

When referring to movement of recently irradiated fuel in the proposed change, the term recently is described in the associated TS Bases, consistent with TSTF-51, as fuel that has occupied part of a critical reactor core within the previous 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months . This time is based on the input assumption in the FHA analysis, which shows that, following this fuel decay period, off-site and MCR doses remain below the 10 CFR 50.67 dose limits without reliance on containment closure, auxiliary building closure, or SFP room filtration from the PPAFES and associated actuation instrumentation.

The operability requirements of the Technical Specifications specified herein are modified to reflect that reactor vessel water level, SFP water level, and decay time are the primary success path for mitigating an FHA. The isolation, pressurization, and filtration of the MCR continues to be assumed in the FHA analysis, and therefore, these requirements are not modified by the proposed amendment request.

As specified in Attachment 12 of this Enclosure, SNC will establish administrative controls that ensure the following guidelines specified in Sub-subsection 11.3.6.5 of NUMARC 93-01 (Reference 8), will be included in the assessment of systems removed from service during fuel handling or core alterations:

Ventilation system and radiation monitor availability should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the proposed license amendment is the reduction in doses due to such decay.

The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

A single normal or contingency method to promptly close containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure; rather the prompt methods should enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

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Enclosure to NL-22-0208 Basis for Proposed Change Regarding proposed deletion of TS 3.9.1, Required Action A.1, suspending core alterations has no effect on the initial conditions or mitigation of refueling mode (i.e., Mode 6) design basis accidents or transients, and this requirement applies an operational burden with no corresponding safety benefit. The purpose of maintaining boron concentration within limits in Mode 6 ensures that a core keff of 0.95 is maintained during fuel handling operations. If boron concentration is not within the required limit, the appropriate action is to immediately suspend positive reactivity additions (current Required Action A.2). There are two evolutions encompassed under the term Core Alterations that could negatively affect the shutdown margin; the addition of fuel and the withdrawal of control rods. However, Required Action A.2 (proposed Required Action A.1), requires immediate suspension of positive reactivity changes. The immediate suspension of positive reactivity changes would include both the addition of fuel to the reactor vessel and the withdrawal of control rods. Another accident considered in MODE 6 that could affect shutdown margin is a dilution event. A boron dilution accident is mitigated by stopping the dilution. Additionally, allowing continuation of some core alterations may, in fact, increase core shutdown margin. For example, removal of one or more irradiated fuel assemblies from the core in the proper sequence or inserting a reactivity control component can increase overall shutdown margin. Therefore, prohibiting core alterations in this condition is unnecessary and possibly eliminates an option to restore core shutdown margin.

As a result, the requirement to immediately suspend core alterations when boron concentration is not within the required limit in Mode 6 is deleted.

Regarding the addition of the proposed Note to TS 3.9.3 Required Action A.1: the existing Required Actions A.1 and A.2, which require immediately suspending core alterations and positive reactivity additions, are unchanged. Suspending positive reactivity additions prohibits diluting the boron concentration of the coolant in the RCS, the loading of fuel assemblies or sources into the core, or the removal of reactivity control components. Suspending core alterations also prohibits any movement of fuel, sources, or reactivity control components in the reactor core. A proposed Note to Required Action A.1 permits fuel assemblies, sources, and reactivity control components to be moved, if necessary, to restore an inoperable source range neutron flux monitor. The source range neutron flux monitors are located outside the reactor core in wells in the concrete reactor shield. The radiation levels in these wells can be very high if fuel assemblies are nearby. Troubleshooting, repair, or replacement of the inoperable source range neutron flux monitors may require moving fuel, sources, or reactivity control components away from the source range neutron flux monitor location to minimize the radiation dose to the workers. Also, in accordance with the definition of Core Alterations specified in TS Section 1.1, if movement of a fuel assembly, source, or reactivity control component is in progress when it is discovered that the required source range neutron flux monitor is inoperable, the component may be placed in a safe location. Therefore, in the event one or more source range neutron flux monitors are inoperable, the required actions continue to minimize actions that could result in reactivity changes within the core, while providing the ability to safely restore source range neutron monitoring capability.

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Enclosure to NL-22-0208 Basis for Proposed Change 3.14 Variations from TSTF-51 and TSTF-471 The proposed amendment is based on the STS changes described in TSTF-51, Revision 2, and TSTF-471, Revision 1, but SNC proposes variations from the NUREG-1431 markups in TSTF-51 and TSTF-471, as identified below and include differing TS numbers and TS titles, where applicable.

1. The definition of CORE ALTERATION is being retained in TS Section 1.1, Definitions, because this terminology continues to be used in a number of TSs, which are not being modified as a result of this amendment request.

This is an administrative variation from TSTF-471.

2. The control room emergency filtration system (CREFS) actuation instrumentation and the CREFS continue to be assumed to provide isolation, pressurization, and filtration of the MCR in the event of an FHA. Since this system and associated isolation instrumentation are mitigation systems necessary to maintain dose to personnel in the MCR below the regulatory and regulatory guidance limits for an FHA, the following TSs and support TSs and associated Bases are not modified:

TS 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS) Actuation Instrumentation, TS 3.7.10, Control Room Emergency Filtration/Pressurization System (CREFS) - Both Units Operating, TS 3.7.11, Control Room Emergency Filtration/Pressurization System (CREFS) - One Unit Operating, TS 3.8.2, AC Sources - Shutdown, TS 3.8.5, DC Sources - Shutdown, TS 3.8.8, Inverters - Shutdown, TS 3.8.10, Distribution Systems - Shutdown, and TS 3.9.7, Refueling Cavity Water Level.

This is a plant-specific variation from TSTF-51 and 471.

3. The applicability requirements associated with the containment ventilation isolation instrumentation are shown in TS Table 3.3.6-1. This is a presentation difference from the applicability requirements shown in the NUREG-1431 TS 3.3.6 marked up pages in TSTF-51. However, the proposed changes to footnote (c) in TS Table 3.3.6-1 are consistent with those shown in TSTF-51. These proposed changes are administrative variations from TSTF-51.

4. TS 3.9.3, Nuclear Instrumentation, Required Actions were not modified in accordance with TSTF-471. However, proposed Note added to Required Action A.1 is consistent with the intent of the proposed Note in TSTF-571-T, Revise Actions for Inoperable Source Range Neutron Flux Monitor (Reference 6). TSTF-571-T was accepted for use by the NRC as documented in a letter to the TSTF dated October 4, 2018 (Reference 7).

Movement of fuel sources and reactivity control components within the reactor vessel is currently covered by the Core Alteration definition. Since E-22

Enclosure to NL-22-0208 Basis for Proposed Change the VEGP TSs retain the definition of Core Alteration, the required action continues to require suspension of core alterations and the note was modified to use the term Core Alterations. These proposed changes are considered administrative variations from TSTF-471.

SNC considers the differences from TSTF-51 and TSTF-471 listed herein to be either: 1) necessary variations to maintain the requirements for required safety systems assumed in the VEGP FHA analysis; or 2) minor variations or deviations that are administrative in nature.

TSTF-490 3.15 TSTF-490 SNC has reviewed References 2 and 3. SNC has also applied the methodology in Reference 1 to develop the proposed TS changes. SNC has also concluded that the justifications presented in TSTF-490, Revision 0, and the model SE prepared by the NRC staff are applicable to VEGP, Units 1 and 2, and justify this amendment for the incorporation of the changes to the VEGP TS.

3.15.1 Variations SNC is not proposing any variations or deviations from the applicable parts of the NRC staffs model safety evaluation. SNC is proposing the following variations from the TS changes described in the TSTF-490, Revision 0.

The VEGP TS include a Surveillance Frequency Control Program.

Therefore, the periodic Surveillance Frequencies shown in TSTF-490 are replaced with the statement, "In accordance with the Surveillance Frequency Control Program."

Additionally, SNC is proposing an administrative change which deletes the global NOTE regarding LCO 3.0.4c applicability. This NOTE is replaced by the same NOTE added to Action A (administrative change) and to Action B (TSTF-490). The NOTE does not apply to Action C (shutdown action).

3.15.2 Calculation of Dose Equivalent Xenon-133 TSTF-490 Revision 0 (described in USNRC ADAMS ML052630462, ML070250176, ML16113A402, and ML18256A027) defines Dose Equivalent Xenon-133 (DE Xe-133) as that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of the Noble Gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. The dose consequences of a given radionuclide is proportional to the product of its RCS concentration and its dose conversion factor (DCF):

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Enclosure to NL-22-0208 Basis for Proposed Change Dose Consequences [Concentration] * [Dose Conversion Factor]

The definition of DE Xe-133 can be developed as follows:

[DE Xe-133 (Ci/gm)] * [DCFXe-133] = [CNOM-i

DCFi]

DE Xe-133 (Ci/gm) = {[C

DCF ]}/DCF i i Xe-133 where:

CNOM-i = Nominal RCS concentration of individual noble gas isotope i (Ci/g) based on clad defects in 1% of the fuel rods DCFi = DCF for individual isotope i [(REM/sec)/(µCi/m3)]

DCFXe-133 = DCF for Xe-133 [(REM/sec)/(µCi/m3)]

The determination of Dose Equivalent Xe-133 is performed using the effective DCFs for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

Since the nominal Noble Gas concentrations based upon 1% cladding defects may differ from their Dose Equivalent Xe-133 concentrations, the nominal concentrations are adjusted to meet the RCS Dose Equivalent Xe-133 specific activity limit as follows:

Dose Equivalent Xe-133 Noble Gas Concentration for Noble = Adjustment Factor

Isotope i Nominal Gas Isotope i Concentration Dose Equivalent Xe-133 Limit CDEX-i (µCi/g) = (µCi/g)

C NOM-i (µCi/g)

{ [CNOM-i

DCFi]}/DCFXe-133 The Dose Equivalent Xe-133 Limit per TSTF-490 is 280 µCi/g. The calculation of the individual noble gases dose equivalent Xe-133 concentrations are shown in the table below. The fuel design margin applied to the core source term is not applied here because as is apparent from the above equation for CDEX-i, it will cancel out.

Nuclide CNOM-i DCFi CNOM-i

DCFi CDEX-i (Ci/g) (REM/sec)/ (REM/sec)/ (Ci/g)

(Ci/m^3) (g/m^3)

Kr-85 9.03E+00 4.40E-10 3.98E-09 3.84E+00 Kr-85m 1.75E+00 2.77E-08 4.85E-08 7.46E-01 E-24

Enclosure to NL-22-0208 Basis for Proposed Change Nuclide CNOM-i DCFi CNOM-i

DCFi CDEX-i (Ci/g) (REM/sec)/ (REM/sec)/ (Ci/g)

(Ci/m^3) (g/m^3)

Kr-87 1.15E+00 1.52E-07 1.75E-07 4.87E-01 Kr-88 3.85E+00 3.77E-07 1.45E-06 1.64E+00 Xe-131m 3.43E+00 1.44E-09 4.93E-09 1.46E+00 Xe-133 2.70E+02 5.77E-09 1.56E-06 1.15E+02 Xe-133m 3.75E+00 5.07E-09 1.90E-08 1.59E+00 Xe-135 8.16E+00 4.40E-08 3.59E-07 3.47E+00 Xe-135m 5.15E-01 7.55E-08 3.88E-08 2.19E-01 Xe-138 6.50E-01 2.13E-07 1.39E-07 2.77E-01 CNOM-i = Nominal RCS Noble Gas concentration for isotope i (µCi/g)

DCFi = Effective DCF for isotope i, Table III.1, FGR-12 (REM/sec)/(µCi/m3)

DE Xe-133 = Dose Equivalent Xe-133 concentration (µCi/g)

DE Xe-133 = [ (CNOM-i

DCFi]/DCFXe-133 (CNOM

DCF]i = 3.80E-06 (REM/sec)/(g/m3)

DCFXe-133 = 5.77E-09 (REM/sec)/(Ci/m3)

DE Xe-133 = 659 µCi/g 133 280 /"

= = = 0.425 133 659 /"

CDEX-i = 0.425

CNOM-i 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria Title 10 Code of Federal Regulations Section 50.36, Technical Specifications Changes to the VEGP TSs are proposed for the adoption of TSTF-51, TSTF-471, and TSTF-490. A description of these proposed changes and their relationship to applicable regulatory requirements and guidance are provided herein.

Title 10 Code of Federal Regulations Section 50.67, Accident Source Term On December 23, 1999, the NRC published 10 CFR 50.67, Accident Source Term, in the Federal Register. This regulation provides a mechanism for licensed power reactors to replace the current accident source term used in the DBA analysis with an AST. The direction provided in 10 CFR 50.67 is that E-25

Enclosure to NL-22-0208 Basis for Proposed Change licenses who seek to revise their current accident source term in design basis radiological consequence analyses shall apply for a LAR under 10 CFR 50.90.

4.1.1 Additional Applicable Regulatory Criteria for TSTF-51 and TSTF-471 The TSs satisfy 10 CFR 50.36, Technical specifications. The following systems and parameters meet one or more of the criteria of 10 CFR 50.36(c)(2)(ii):

Containment and associated containment purge and exhaust isolation instrumentation, Piping Penetration Area Filtration and Exhaust System (PPAFES) and associated actuation instrumentation, Boron concentration requirement in Mode 6, and Neutron instrumentation requirements in Mode 6.

The proposed amendment revises the TS applicability of these systems and parameters to eliminate the requirements during core alterations, and during movement of irradiated fuel assemblies that have decayed beyond the decay period assumed in the VEGP FHA analysis because these requirements are no longer assumed in the mitigation of an FHA or the potential radioactive release as a result of dropping of a non-irradiated fuel assembly, source, or reactivity control component onto the reactor core during core alterations. The proposed amendment does not alter requirements associated with the CREFS and associated instrumentation, which are assumed to mitigate the effects of a radiological release to MCR personnel due to an FHA, and continues to maintain requirements associated with structures, systems, and components that are part of the primary success path and actuate to mitigate the related design basis accidents and transients. The proposed amendment continues to provide appropriate remedial actions and shutdown requirements required by 10 CFR 50.36(c)(2)(i) for any system requiring an LCO pursuant the criteria of 10 CFR 50.36(c)(2)(ii).

10 CFR 50.67, Accident source term - Note that this License Amendment request includes conversion of the VEGP source terms to the ASTs consistent with 10 CFR 50.67. Thus, the modified The VEGP FHA analysis of record meets the requirements of 10 CFR 50.67.

Accident source terms have not been modified as a result of the proposed amendment. SNC has determined that the inputs and assumptions related to atmospheric dispersion related to the FHA analysis are not changed as a result of the proposed license amendment. Therefore, the revised VEGP FHA analysis continues to meet the requirements of 10 CFR 50.67.

In addition, the following 10 CFR Part 50, Appendix A General Design Criteria (GDCs) are related to the proposed change:

GDC 13: Instrumentation and control. The proposed amendment does not alter the design of the applicable instrumentation that monitor E-26

Enclosure to NL-22-0208 Basis for Proposed Change variables and systems over their anticipated ranges for normal operation for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety.

GDC 16: Containment design. The proposed amendment does not alter the containment design or the associated systems design. The containment and associated systems, when required, will continue to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment as previously licensed and approved by the NRC. During movement of recently irradiated fuel assemblies when containment integrity is relaxed, SNC, as previously committed, will continue to establish and implement administrative controls that ensure that the open personnel airlock and any open penetrations can and will be promptly closed, following containment evacuation, in the event of an FHA inside containment.

GDC 19: Control room. The proposed amendment does not alter the design or operation of the control room envelope or the CREFS. To support the proposed amendment, the input assumptions have been revised in the FHA analysis. However, FHA analysis results show that the radiological dose to the MCR personnel continues to be within the requirements of GDC-19 as updated for consistency with the TEDE criterion in 10 CFR 50.67.b.2.iii. Adequate radiation protection continues to be provided permitting access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

GDC 20: Protection system functions. The proposed amendment does not alter the design of reactivity control protection systems or instrumentation that sense accident conditions to initiate systems or components important to safety. The change relaxes the requirements for instrumentation of systems not assumed in the mitigation of an FHA.

GDC 21: Protection system reliability and testability. The proposed amendment does not alter the design of any protection system, including the containment ventilation instrumentation and the CREFS actuation instrumentation. Therefore, the protection system design continues to provide high functional reliability and inservice testability commensurate with the safety functions to be performed and continues to be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy. The containment ventilation instrumentation and the CREFS actuation instrumentation designs continue to permit periodic testing of its functioning when the reactor is in operation as previously licensed and approved by the NRC.

GDC 22: Protection system independence. The proposed amendment does not alter the design of any protection system, including the containment ventilation instrumentation and the CREFS actuation instrumentation. Therefore, the protection system design continues to E-27

Enclosure to NL-22-0208 Basis for Proposed Change assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function to the extent previously licensed and approved by the NRC.

GDC 23: Protection system failure modes. The proposed amendment does not alter the design of any protection system, including the containment ventilation instrumentation and the CREFS actuation instrumentation. Therefore, the protection system design continues to fail into a safe state or into a state demonstrated to be acceptable as previously licensed and approved by the NRC.

GDC 24: Separation of protection and control systems. The proposed amendment does not alter the design of any protection system, including the containment ventilation instrumentation and the CREFS actuation instrumentation. Therefore, the protection system design continues to be separated from control systems as previously licensed and approved by the NRC.

GDC 64: Monitoring radioactivity releases. The proposed amendment does not alter the design of any radioactivity monitoring instrumentation, including the containment ventilation instrumentation and the CREFS actuation instrumentation. Means continue to be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents (e.g., FHA).

4.1.2 Additional Applicable Regulatory Criteria for TSTF-490 A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on March 19, 2017 (Reference 2), the NRC Notice for Comment published on November 20, 2006 (Reference 3), and TSTF-490, Revision 0.

4.2 Precedent 4.2.1 A number of AST submittals have been reviewed and approved by the NRC since RG 1.183 was published and have helped to inform the content of this application, specifically Joseph M. Farley Nuclear Plant - Units 1 and 2, License Amendments 216 and 213, respectively (NRC ADAMS Accession No. ML17271A265).

4.2.2 STS Traveler TSTF-51 was approved by the NRC staff and incorporated into the STS NUREGs, Revision 2, published in June 2001, which was also approved by the NRC staff. A number of facilities have adopted, as technically practical, TSTF-51. For example: Joseph M. Farley Nuclear Plant, Units 1 and 2, License Amendments 223 and 220, respectively (NRC ADAMS Accession No. ML19071A138), Indian Point Nuclear E-28

Enclosure to NL-22-0208 Basis for Proposed Change Generating Unit 2, License Amendment 238 (NRC ADAMS Accession Nos. ML033160528 and ML033210260), North Anna Power Station, Units 1 and 2, License Amendments 231 and 212, respectively (NRC ADAMS Accession Nos. ML021200265, ML021220108, and ML021220166),

Beaver Valley Units 1 and 2, License Amendments 278 and 161, respectively (NRC ADAMS Accession Nos. ML070160593 and ML070390284), Watts Bar Nuclear Plant, Unit 1, License Amendment 35 (NRC ADAMS Accession Nos. ML020100062 and ML020280264), and Byron Units 1 and 2, License Amendments 147 and Braidwood Units 1 and 2, License Amendments 140 (NRC ADAMS Accession No. ML062340420).

4.2.3 STS Traveler TSTF-471 was approved for use by the NRC staff and incorporated into the applicable STS NUREGs, Revision 4, published in April 2012. Some facilities have adopted, as technically practical, TSTF-471. For example: Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Amendments 279 and 256, respectively (NRC ADAMS Accession Nos.

ML062350447 and ML062690054).

4.2.4 STS Traveler TSTF-490 was approved for use by the NRC staff and incorporated into the applicable STS NUREGs, Revision 5, published in September 2021. Some facilities have adopted as technically practical, TSTF-490. For example: Salem Generating Station, Units 1 and 2, Amendments 337 and 318, respectively (NRC ADAMS Accession No. ML21110A052).

4.3 Significant Hazards Consideration Southern Nuclear Operating Company (SNC) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The proposed amendment revises certain Technical Specifications (TSs) to remove the requirements for engineered safety feature (ESF) systems to be operable after sufficient radioactive decay of irradiated fuel has occurred following a plant shutdown. Following sufficient radioactive decay, these systems are no longer required during a fuel handling accident (FHA) to ensure main control room (MCR) personnel dose remains below the 10 CFR 50.67(b)(2)(iii) dose limit and off-site dose remains below the accident dose limit specified in the NRC standard review plan (SRP), which represent a small fraction of the 10 CFR 50.67 dose limits.

The proposed amendment also revises certain TS actions that are not needed to mitigate accidents postulated during shutdown. Specifically, the requirement to immediately suspend core alterations when boron concentration is not within the required limit in refueling condition is deleted. In addition, when one or more required source range neutron flux monitors are inoperable in the refueling condition, a note added to the actions will permit fuel assemblies, sources, and reactivity control components to be moved if necessary to restore an inoperable source range neutron flux monitor to operable status.

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Enclosure to NL-22-0208 Basis for Proposed Change Furthermore, SNC has reviewed the proposed no significant hazards consideration determination for TSTF-490 published in the Federal Register on March 19, 2007 (72 FR 12838) as part of the Consolidated Line Item Improvement Process (CLIIP). SNC has concluded that the proposed determination presented in the notice is applicable to VEGP and the determination is hereby incorporated by reference.

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No There are no physical changes to the plant being introduced by the proposed changes to the accident source term. Implementation of Alternative Source Term (AST) and the new atmospheric dispersion factors have no impact on the probability for initiation of any Design Basis Accidents (DBAs). Once the occurrence of an accident has been postulated, the new accident source term and atmospheric dispersion factors are an input to analyses that evaluate the radiological consequences. The proposed changes do not involve a revision to the design or manner in which the facility is operated that could increase the probability of an accident previously evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR).

Based on the AST analyses, there are no proposed changes to performance requirements and no proposed revision to the parameters or conditions that could contribute to the consequences of an accident previously discussed in Chapter 15 of the FSAR. Plant-specific radiological analyses have been performed using the AST methodology and new atmospheric dispersion factors (X/Qs) have been established. Based on the results of these analyses, it has been demonstrated that the Control Room and off-site dose consequences of the limiting events considered in the analyses meet the regulatory guidance provided for use with the AST, and the doses are within the limits established by 10 CFR 50.67.

The proposed amendment does not involve a physical change to the containment or spent fuel area systems, nor does it change the safety function of the containment or associated instrumentation. The subject ESF systems are not assumed in the mitigation of an FHA after sufficient radioactive decay of irradiated fuel has occurred. The revised FHA dose analysis shows that MCR dose remains below the 10 CFR 50.67(b)(2)(iii) dose limit and off-site dose remains below the accident dose limit specified in the NRC SRP, which represents a small fraction of the 10 CFR 50.67 dose limits.

E-30

Enclosure to NL-22-0208 Basis for Proposed Change Elimination of the action to suspend core alterations in the event boron concentration is not within the required limit in refueling condition does not alter the initiation or consequences of a boron dilution event and the required actions continue to prohibit positive reactivity additions until reactor core shutdown margin can be restored to within the required limit.

Permitting fuel assemblies, sources, and reactivity control components to be moved to restore an inoperable source range neutron flux monitor to operable status when one or more required source range neutron flux monitors are inoperable does not significantly alter the probability or consequences of any previously evaluated refueling accident or transient. The required actions continue to minimize actions that could result in reactivity changes within the core, while providing the ability to safely restore source range neutron monitoring capability.

Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or the consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Implementation of the AST and the associated Technical Specification changes and new X/Qs have no impact to the initiation of any DBAs. These changes do not affect the design function of modes of operation of structures, systems and components in the facility prior to a postulated accident. Since structures, systems and components are operated no differently after the AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents.

With respect to a new or different kind of accident, there are no proposed design changes to the safety related plant structures, systems, and components (SSCs); nor are there any changes in the method by which safety related plant SSCs perform their specified safety functions. The proposed amendment will not affect the normal method of plant operation or revise any operating parameters. No new accident scenarios, transient precursor, failure mechanisms, or limiting single failures will be introduced as a result of this proposed change and the failure modes and effects analyses of SSCs important to safety are not altered as a result of this proposed change. The proposed amendment does not alter the design or performance of the related SSCs, and, therefore, does not constitute a new type of test.

No changes are being proposed to the procedures that operate the plant equipment and the change does not have a detrimental impact on the E-31

Enclosure to NL-22-0208 Basis for Proposed Change manner in which plant equipment operates or responds to an actuation signal.

Consequently, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The AST analyses have been performed using approved methodologies to ensure that analyzed events are bounding and safety margin has not been reduced. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67. Thus, by meeting the applicable regulatory limits for AST, there is no significant reduction in a margin of safety.

The proposed amendment does not involve a physical change to the containment, nor does it change the safety function of the containment or associated instrumentation. The subject ESF systems are not assumed in the mitigation of an FHA after sufficient radioactive decay of irradiated fuel has occurred. The revised VEGP FHA dose analysis shows that MCR dose remains below the 10 CFR 50.67(b)(2)(iii) dose limit and off-site dose remains below the accident dose limit specified in the NRC SRP, which represents a small fraction of the 10 CFR 50.67 dose limits.

Elimination of the action to suspend core alterations does not reduce the margin of safety in the event boron concentration is not within the required limit in refueling condition because the remaining required actions continue to prohibit positive reactivity additions until reactor core shutdown margin can be restored to within the required limit.

Permitting fuel assemblies, sources, and reactivity control components to be moved to restore an inoperable source range neutron flux monitor to operable status when one or more required source range neutron flux monitors are inoperable does not significantly reduce the margin of safety. The required actions continue to minimize actions that could result in reactivity changes within the core, while providing the ability to safely restore source range neutron monitoring capability.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the proposed amendment does not involve a significant reduction in margin of safety.

4.4 Conclusion In conclusion, based on the considerations discussed above, SNC concludes:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities E-32

Enclosure to NL-22-0208 Basis for Proposed Change will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 Environmental Consideration The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Furthermore, in regard to the implementation of TSTF-490, SNC has reviewed the environmental consideration included in the model SE published in the Federal Register on March 19, 2007 (72 FR 12838) as part of the CLIIP. SNC has concluded that the staffs findings presented therein are applicable to VEGP and the determination is hereby incorporated by reference for this application.

6.0 References

1. Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors"

2. Federal Register 71 FR 12838, Notice of Availability of Model Application Concerning Technical Specification (TS) Improvement Regarding TSTF-490, Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity TS Using Consolidated Line Item Improvement Process, dated March 19, 2007 (NRC ADAMS Accession No. ML070250176)

3. Federal Register 71 FR 67170, Notice of Opportunity to Comment on Model Safety Evaluation and Model License Amendment Request on Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process, dated November 20, 2006

4. Regulatory Guide (RG) 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants

5. STS Change Traveler TSTF-286, Define Operations Involving Positive Reactivity Additions, Revision 2, dates April 3, 2000.

E-33

Enclosure to NL-22-0208 Basis for Proposed Change

6. Letter from Technical Specification Task Force to NRC, TSTF Input to Lifting the Suspension of Acceptance of Amendment Requests to Adopt TSTF-51, TSTF-471, and TSTF-286, dated August 9. 2018 (NRC ADAMS Accession No. ML18221A561).

7. Letter from V.G. Cusumano (NRC) to Technical Specification Task Force, Plant Specific Adoption of Travelers TSTF-51, Revision 2, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations, TSTF-471, Revision 1, Eliminate Use of Term Core Alterations in Actions and Notes, and TSTF-286, Revision 2, Operations Involving Positive Reactivity Additions, dated October 4, 2018 (NRC ADAMS Accession No. ML17346A587).

8. Nuclear Energy Institute NUMARC 93-01, Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4A, April 2011 (NRC ADAMS Accession No. ML11116A198).

E-34 to Enclosure Technical Specification Pages (Markup)

Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 1 Technical Specification Pages (Markup)

A1-1

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (COT) actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in EPA Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, EPA-520/1-88-020, September 1988. DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine (continued)

Vogtle Units 1 and 2 1.1-2 Amendment No. 201 (Unit 1)

Amendment No. 184 (Unit 2)

Definitions 1.1 isotopes I-131, I-132, I-133, I-134, and I-135 actually present.

The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil.

(continued)

Vogtle Units 1 and 2 1.1-3 Amendment No. 201 (Unit 1)

Amendment No. 184 (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

- AVERAGE shall be the average (weighted in proportion to DISINTEGRATION ENERGY the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 14 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE (ESF) RESPONSE interval from when the monitored parameter exceeds its TIME ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

(continued)

Vogtle Units 1 and 2 1.1-4 Amendment No. 213 (Unit 1)

Amendment No. 196 (Unit 2)

Containment Ventilation Isolation Instrumentation 3.3.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ------------NOTE-------------- C.1 Place and maintain Immediately Only applicable during containment purge and CORE ALTERATIONS or exhaust valves in closed movement of recently position.

irradiated fuel assemblies within containment. OR No radiation monitoring C.2 Enter applicable Immediately channels OPERABLE. Conditions and Required Actions of LCO 3.9.4, OR "Containment Penetrations," for Required Action and containment purge supply associated Completion and exhaust isolation Time for Condition A not penetrations not in met. required status.

Vogtle Units 1 and 2 3.3.6-3 Amendment No. 105 (Unit 1)

Amendment No. 83 (Unit 2)

Containment Ventilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Ventilation Isolation Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CHANNELS REQUIREMENTS TRIP SETPOINT CONDITIONS

1. Manual Initiation 1,2,3,4 2 SR 3.3.6.6 NA

2. Automatic Actuation Logic SR 3.3.6.2 NA and Actuation Relays 1,2,3,4 2 SR 3.3.6.3 SR 3.3.6.5

3. Containment Radiation SR 3.3.6.1 SR 3.3.6.4 1,2,3,4,6(c)

(a) 2 SR 3.3.6.7 SR 3.3.6.8 (b)

a. Gaseous (RE-2565C)

(b)

b. Particulate (RE-2565A)

(b)

c. Iodine (RE-2565B) 15 mr/h(c)

d. Area Low Range 50x background(d)

(RE-0002, RE-0003)

4. Safety Injection(d) 1,2,3,4 Refer to LCO 3.3.2, "ESFAS Instrumentation,"Function 1, for all initiation functions and requirements.

(a) Containment ventilation radiation (RE-2565) is treated as one channel and is considered OPERABLE if the particulate (RE-2565A) and iodine monitors (RE-2565B) are OPERABLE or the noble gas monitor (RE-2565C) is OPERABLE.

(b) Setpoints will not exceed the limits of Specifications 5.5.4.h and 5.5.4.i of the Radioactive Effluent Controls Program.

(c) During CORE ALTERATIONS and movement of recently irradiated fuel assemblies within containment.

(d) During MODES 1, 2, 3, and 4.

Vogtle Units 1 and 2 3.3.6-6 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits. RCS DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1, and 2, 3, and 4 MODE 3 with RCS average temperature (Tavg) 500F.

ACTIONS

NOTE--------------------------------------------------------

LCO 3.0.4c is applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT -------------------NOTE-------------------- Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I-131 not within limit> LCO 3.0.4.c is applicable.

1.0 Ci/gm. ------------------------------------------------

A.1 Verify DOSE EQUIVALENT I-131 60

µCi/gmwithin the acceptable region of Figure 3.4.16-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT I-131 to within limit.

B. Gross specific activity of -------------------NOTE-------------------- 6 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the reactor coolant LCO 3.0.4.c is applicable.

DOSE EQUIVALENT ------------------------------------------------

XE-133 not within limit.

B.1 Be in MODE 3 with Tavg < 500FRestore DOSE EQUIVALENT XE-133 to within limit.

(continued)

Vogtle Units 1 and 2 3.4.16-1 Amendment No. 180 (Unit 1)

Amendment No. 161 (Unit 2)

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Tavg < 500F.

Time of Condition A or B not met. AND OR C.2 Be in MODE 5 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT I-131 > 60 µCI/gmin the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 -------------------------------NOTE-------------------------- In accordance with Only required to be performed in MODE 1. the Surveillance

Frequency Control Program Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity 280 Ci/gm.Verify reactor coolant gross specific activity 100/ Ci/gm.

SR 3.4.16.2 ----------------------------NOTE-----------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity 1.0 Ci/gm. the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER Vogtle Units 1 and 2 3.4.16-2 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

Delete Page in its entirety. RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.3 -----------------------------NOTE----------------------------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine from a sample taken in MODE 1 In accordance with after a minimum of 2 effective full power days and the Surveillance 20 days of MODE 1 operation have elapsed since Frequency Control the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Program Vogtle Units 1 and 2 3.4.16-3 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

RCS Specific Activity 3.4.16 Delete Page in its entirety.

FIGURE 3.4.16-1 REACTOR COOLANT DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 Ci/gram DOSE EQUIVALENT I-131 Vogtle Units 1 and 2 3.4.16-4 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR.

APPLICABILITY: MODE 6.

NOTE-------------------------------------------------

Only applicable to the refueling canal and refueling cavity when connected to the RCS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.

AND A.21 Suspend positive Immediately reactivity additions.

AND A.32 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.9.1-1 Amendment No. 180 (Unit 1)

Amendment No. 161 (Unit 2)

Unborated Water Source Isolation Valves 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Unborated Water Source Isolation Valves LCO 3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position.

NOTE---------------------------------------------

Valves in the flowpath from the RMWST, through the chemical mixing tank, to the suction of the charging pumps may be opened under administrative control provided the reactor coolant system is in compliance with Specification 3.9.1 and the high flux at shutdown alarm is OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS

NOTE----------------------------------------------------------------

Separate Condition entry is allowed for each unborated water source isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE-------------- A.1 Suspend CORE Immediately Required Action A.3 2 ALTERATIONS.

must be completed whenever Condition A is AND entered.

A.21 Initiate actions to secure Immediately valve in closed position.

One or more valves not secured in closed position. AND A.32 Perform SR 3.9.1.1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (verify boron concentration).

Vogtle Units 1 and 2 3.9.2-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Nuclear Instrumentation LCO 3.9.3 Two source range neutron flux monitors shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One source range neutron A.1 ----------NOTE------------- Immediately flux monitor inoperable. CORE ALTERATIONS may continue to restore an inoperable source range neutron flux monitor.

A.1 Suspend CORE ALTERATIONS.

AND A.2 Suspend positive Immediately reactivity additions.

B. --------------NOTE------------- B.1 Initiate action to restore Immediately Condition A entry is one source range required when Condition B neutron flux monitor to is entered. OPERABLE status.

AND Two source range neutron flux monitors inoperable. B.2 Perform SR 3.9.1.1 Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (verify boron concentration).

Vogtle Units 1 and 2 3.9.3-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4 The containment penetrations shall be in the following status:

a. The equipment hatch is capable of being closed and held in place by four bolts;

b. The emergency and personnel air locks are isolated by at least one air lock door, or if open, the emergency and personnel air locks are isolable by at least one air lock door with a designated individual available to close the open air lock door(s); and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by at least two OPERABLE Containment Ventilation Isolation valves

NOTE---------------------------------------------

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTERATIONS, During movement of recently irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

AND A.21 Suspend movement of Immediately recently irradiated fuel assemblies within containment.

Vogtle Units 1 and 2 3.9.4-1 Amendment No. 181 (Unit 1)

Amendment No. 162 (Unit 2) to Enclosure Technical Specification Pages (Clean Typed Pages)

Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 2 Technical Specification (Clean Typed Pages)

A2-1

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (COT) actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present.

The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.

(continued)

Vogtle Units 1 and 2 1.1-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE (ESF) RESPONSE interval from when the monitored parameter exceeds its TIME ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

(continued)

Vogtle Units 1 and 2 1.1-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

Vogtle Units 1 and 2 1.1-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

OPERABLE OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:)

a. Described in Chapter 14 of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, Cold Overpressure Protection System (COPS) arming temperature and the nominal PORV setpoints for the COPS, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. Unit operation within these operating limits is addressed in individual specifications.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper RATIO (QPTR) excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3625.6 MWt.

(continued)

Vogtle Units 1 and 2 1.1-5 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval SYSTEM (RTS) RESPONSE from when the monitored parameter exceeds its RTS TIME trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay.

The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

Vogtle Units 1 and 2 1.1-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 1.1 Definitions (continued)

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of required alarm, (TADOT) interlock, and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

Vogtle Units 1 and 2 1.1-7 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES

% RATED AVERAGE REACTIVITY THERMAL MODE TITLE REACTOR COOLANT CONDITION (keff) POWER(a) TEMPERATURE (F) 1 Power Operation 0.99 >5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 350 4 Hot Shutdown(b) < 0.99 NA 350 > Tavg > 200 5 Cold Shutdown(b) < 0.99 NA 200 6 Refueling(c) NA NA NA (a) Excluding decay heat.

(b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned.

Vogtle Units 1 and 2 1.1-8 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Containment Ventilation Isolation Instrumentation 3.3.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ------------NOTE-------------- C.1 Place and maintain Immediately Only applicable during containment purge and movement of recently exhaust valves in closed irradiated fuel assemblies position.

within containment.

OR No radiation monitoring channels OPERABLE. C.2 Enter applicable Immediately Conditions and Required OR Actions of LCO 3.9.4, "Containment Required Action and Penetrations," for associated Completion containment purge supply Time for Condition A not and exhaust isolation met. penetrations not in required status.

Vogtle Units 1 and 2 3.3.6-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Containment Ventilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Ventilation Isolation Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CHANNELS REQUIREMENTS TRIP SETPOINT CONDITIONS

1. Manual Initiation 1,2,3,4 2 SR 3.3.6.6 NA

2. Automatic Actuation Logic SR 3.3.6.2 NA and Actuation Relays 1,2,3,4 2 SR 3.3.6.3 SR 3.3.6.5

3. Containment Radiation SR 3.3.6.1 SR 3.3.6.4 1,2,3,4,6(c)

(a) 2 SR 3.3.6.7 SR 3.3.6.8 (b)

a. Gaseous (RE-2565C)

(b)

b. Particulate (RE-2565A)

(b)

c. Iodine (RE-2565B) 15 mr/h(c)

d. Area Low Range 50x background(d)

(RE-0002, RE-0003)

4. Safety Injection(d) 1,2,3,4 Refer to LCO 3.3.2, "ESFAS Instrumentation,"Function 1, for all initiation functions and requirements.

(a) Containment ventilation radiation (RE-2565) is treated as one channel and is considered OPERABLE if the particulate (RE-2565A) and iodine monitors (RE-2565B) are OPERABLE or the noble gas monitor (RE-2565C) is OPERABLE.

(b) Setpoints will not exceed the limits of Specifications 5.5.4.h and 5.5.4.i of the Radioactive Effluent Controls Program.

(c) During movement of recently irradiated fuel assemblies within containment.

(d) During MODES 1, 2, 3, and 4.

Vogtle Units 1 and 2 3.3.6-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT -------------------NOTE-------------------- Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I-131 not within limit. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT I-131 60

µCi/gm.

AND A.2 Restore DOSE EQUIVALENT I-131 to within limit. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B. DOSE EQUIVALENT -------------------NOTE-------------------- 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months XE-133 not within limit. LCO 3.0.4.c is applicable.

B.1 Restore DOSE EQUIVALENT XE-133 to within limit.

(continued)

Vogtle Units 1 and 2 3.4.16-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5 OR 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT I-131 > 60 µCI/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 -------------------------------NOTE-------------------------- In accordance with Only required to be performed in MODE 1. the Surveillance

Frequency Control Program Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity 280 Ci/gm.

SR 3.4.16.2 ----------------------------NOTE-----------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity 1.0 Ci/gm. the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period Vogtle Units 1 and 2 3.4.16-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR.

APPLICABILITY: MODE 6.

NOTE-------------------------------------------------

Only applicable to the refueling canal and refueling cavity when connected to the RCS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend positive Immediately within limit. reactivity additions.

AND A.2 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.9.1-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Unborated Water Source Isolation Valves 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Unborated Water Source Isolation Valves LCO 3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position.

NOTE---------------------------------------------

Valves in the flowpath from the RMWST, through the chemical mixing tank, to the suction of the charging pumps may be opened under administrative control provided the reactor coolant system is in compliance with Specification 3.9.1 and the high flux at shutdown alarm is OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS

NOTE----------------------------------------------------------------

Separate Condition entry is allowed for each unborated water source isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE-------------- A.1 Initiate actions to secure Immediately Required Action A.2 must valve in closed position.

be completed whenever Condition A is entered. AND A.2 Perform SR 3.9.1.1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more valves not (verify boron secured in closed position. concentration).

Vogtle Units 1 and 2 3.9.2-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Nuclear Instrumentation LCO 3.9.3 Two source range neutron flux monitors shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One source range neutron A.1 ----------NOTE------------- Immediately flux monitor inoperable. CORE ALTERATIONS may continue to restore an inoperable source range neutron flux monitor.

Suspend CORE ALTERATIONS.

AND A.2 Suspend positive Immediately reactivity additions.

B. --------------NOTE------------- B.1 Initiate action to restore Immediately Condition A entry is one source range required when Condition B neutron flux monitor to is entered. OPERABLE status.

AND Two source range neutron flux monitors inoperable. B.2 Perform SR 3.9.1.1 Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (verify boron concentration).

Vogtle Units 1 and 2 3.9.3-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4 The containment penetrations shall be in the following status:

a. The equipment hatch is capable of being closed and held in place by four bolts;

b. The emergency and personnel air locks are isolated by at least one air lock door, or if open, the emergency and personnel air locks are isolable by at least one air lock door with a designated individual available to close the open air lock door(s); and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by at least two OPERABLE Containment Ventilation Isolation valves

NOTE---------------------------------------------

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During movement of recently irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend movement of Immediately penetrations not in recently irradiated fuel required status. assemblies within containment.

Vogtle Units 1 and 2 3.9.4-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) to Enclosure Technical Specification Bases Pages (Markup) (For Information Only)

Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 3 Technical Specification Bases Pages (Markup) (For Information Only)

A3-1

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB.

If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 10050.67, "Reactor Site CriteriaAccident Source Term" (Ref. 4).

Vogtle Units 1 and 2 B 2.1.2-1 Revision No. 0

RCS Pressure SL B 2.1.2 BASES SAFETY LIMITS Code,Section III, is 110% of design pressure. Therefore, the SL (continued) on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

SAFETY LIMIT VIOLATIONS If the RCS pressure SL 2.2.2 is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 10050.67, "Reactor Site CriteriaAccident Source Term," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

(continued)

Vogtle Units 1 and 2 B 2.1.2-4 REVISION 16

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT VIOLATIONS (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWB-5000.

4. 10 CFR 10050.67.

5. FSAR, Section 7.2.

Vogtle Units 1 and 2 B 2.1.2-6 REVISION 16

SDM B 3.1.1 BASES APPLICABLE SDM satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii). Even SAFETY ANALYSES though it is not directly observed from the control room, (continued) SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 10050.67, "Reactor Site CriteriaAccident Source Term," limits (Ref. 4). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable. The required SDM is specified in the COLR.

APPLICABILITY In MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits."

ACTIONS The ACTIONS table is modified by a Note prohibiting transition to a lower MODE within the Applicability. LCO 3.0.4 already prohibits entry into MODE 5 from MODE 6, MODE 4 from MODE 5 and into MODE 3 from MODE 4 when SDM requirements are not met.

(continued)

Vogtle Units 1 and 2 B 3.1.1-4 Rev. 3 - 6/05

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Subsection 15.4.9.

3. FSAR, Subsection 15.4.6.

4. 10 CFR 10050.67.

Vogtle Units 1 and 2 B 3.1.1-6 REVISION 14

RTS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Trip System (RTS) Instrumentation BASES BACKGROUND The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs) and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents.

The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RTS, as well as specifying LCOs on other reactor system parameters and equipment performance.

The LSSS, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to occur one or more times during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling (DNB);

2. Fuel centerline melt shall not occur; and

3. The RCS pressure SL of 2750 psia shall not be exceeded.

Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"

also maintains the above values and assures that offsite dose will be within the 10 CFR 20 and 10 CFR 100 50.67 criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 10050.67 (continued)

Vogtle Units 1 and 2 B 3.3.1-1 Rev. 1 - 9/03

ESFAS Instrumentation B 3.3.2 BASES BACKGROUND the "as-found" value of a protection channel setting during a (continued) surveillance. This would result in Technical Specification compliance problems, as well as reports and corrective actions required by the rule which are not necessary to ensure safety. For example, an automatic protection channel with a setting that has been found to be different from the NTSP due to some drift of the setting may still be OPERABLE since drift is to be expected. This expected drift would have been specifically accounted for in the setpoint methodology for calculating the NTSP and thus the automatic protective action would still have ensured that the SL would not be exceeded with the "as-found" setting of the protection channel. Therefore, the channel would still be OPERABLE since it would have performed its safety function and the only corrective action required would be to reset the channel within the established as-left tolerance around the NTSP to account for further drift during the next surveillance interval.

During AOOs, which are those events expected to occur one or more times during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the SL value to prevent departure from nucleate boiling (DNB),

2. Fuel centerline melt shall not occur, and

3. The RCS pressure SL of 2750 psia shall not be exceeded.

Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"

also maintains the above values and assures that offsite dose will be within the 10 CFR 50 and 10 CFR 100 50.67 criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 50.67 limits. Different accident categories are allowed a different fraction of these limits, based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

The ESFAS instrumentation is segmented into four distinct but interconnected modules as identified below:

Field transmitters or process sensors and instrumentation:

provide a measurable electronic signal based on the physical characteristics of the parameter being measured; (continued)

Vogtle Units 1 and 2 B 3.3.2-2 REVISION 20

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES (continued)

APPLICABLE The safety analyses assume that the containment remains SAFETY ANALYSES intact with penetrations unnecessary for core cooling isolated early in the event, within approximately 60 seconds. The isolation of the purge supply and exhaust valves has not been analyzed mechanistically in the dose calculations, although its rapid isolation is assumed. The containment purge supply and exhaust isolation radiation monitors act as backup to the SI signal to ensure closing of the purge supply and exhaust valves for events occurring in MODES 1 through 4. Although not credited in the fuel handling accident analysis, Manual manual isolation (using individual valve handswitches) following a radiation alarm is the assumed means for isolating containment in the event of a fuel handling accident during shutdown. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses less than the acceptance criteria in Regulatory Guide 1.183 Revision 0are below 10 CFR 100 (Ref. 1) limits. Due to radioactive decay, containment is only required to isolate during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months ).

The containment ventilation isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO The LCO requirements ensure that the instrumentation necessary to initiate Containment Ventilation Isolation, listed in Table 3.3.6-1, is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate Containment ventilation isolation at any time by using either of two switches in the control room (containment isolation Phase A switches). Either switch actuates both trains. This action will cause actuation of all components in the same manner as any of the automatic actuation signals.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one CIA handswitch and the interconnecting wiring to the actuation logic cabinet.

(continued)

Vogtle Units 1 and 2 B 3.3.6-2 Rev. 2-10/01

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES LCO 2. Automatic Actuation Logic and Actuation Relays (continued)

The LCO requires two channels of Automatic Actuation Logic and Actuation Relays OPERABLE to ensure that no single random failure can prevent automatic actuation.

Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b, SI. The applicable MODES and specified conditions for the Containment ventilation isolation portion of these Functions are different and less restrictive than those for their SI roles. If one or more of the SI Functions becomes inoperable in such a manner that only the Containment Ventilation Isolation Function is affected, the Conditions applicable to their SI Functions need not be entered. The less restrictive Actions specified for inoperability of the Containment Ventilation Isolation Functions specify sufficient compensatory measures for this case.

3. Containment Radiation The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment ventilation isolation remains OPERABLE.

During CORE ALTERATIONS or movement of recently irradiated fuel assemblies in containment, the required channels provide input to control room alarms to ensure prompt operator action to manually close the containment purge and exhaust valves. It is also acceptable during CORE ALTERATIONS or movement of recently irradiated fuel to meet the requirements of this LCO by maintaining the radiation monitoring instrumentation necessary to initiate containment ventilation isolation OPERABLE, in accordance with the requirements stated for MODES 1, 2, 3, and 4 operability. The purge exhaust radiation detectors (RE-2565A, B&C) are treated as one channel which is considered OPERABLE if the particulate (RE-2565A) and iodine (RE-2565B) monitors are OPERABLE or the noble gas monitor (RE-2565C) is OPERABLE.

In addition, two individual channels of containment area low range gamma monitors (RE-0002 & RE-0003) are provided. The two required radiation monitoring channels may be made up of any combination of the above described channels.

(continued)

Vogtle Units 1 and 2 B 3.3.6-4 Rev. 1-3/99

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES LCO 4. Safety Injection (continued)

Refer to LCO 3.3.2, Function 1, for all initiating Functions and requirements. The safety injection initiation function is applicable in MODES 1, 2, 3, and 4 only.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, Containment Radiation, and Safety Injection Functions are required OPERABLE in MODES 1, 2, 3, and 4. Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment. Therefore, the Containment ventilation isolation instrumentation must be OPERABLE in these MODES.

During CORE ALTERATIONS or movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months ) in containment, the air locks may be open provided they are isolable per LCO 3.9.4. Since the air locks can only be closed manually, it is assumed that containment ventilation isolation is accomplished by manually closing the purge and exhaust ventilation valves. Therefore, only OPERABLE radiation monitors are required to alert the operators of the need for containment ventilation isolation.

While in MODES 5 and 6 without fuel handling in progress, the containment ventilation isolation instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference 1.

ACTIONS The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by unit specific calibration procedures. Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. This determination is generally made during the performance of a COT, when the process instrumentation is set up for adjustment to bring it within specification. If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1.

The Completion Time(s) of (continued)

Vogtle Units 1 and 2 B 3.3.6-6 Rev. 1-3/99

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES ACTIONS C.1 and C.2 (continued)

Required Action A.1. If no radiation monitoring channels are operable or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action to place and maintain containment purge supply and exhaust isolation valves in their closed position is met or the applicable Conditions of LCO 3.9.4, "Containment Penetrations," are met for each penetration not in the required status. The Completion Time for these Required Actions is Immediately.

A Note states that Condition C is applicable during CORE ALTERATIONS and during movement of recently irradiated fuel assemblies within containment.

SURVEILLANCE A Note has been added to the SR Table to clarify that REQUIREMENTS Table 3.3.6-1 determines which SRs apply to which Containment Ventilation Isolation Functions.

SR 3.3.6.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is (continued)

Vogtle Units 1 and 2 B 3.3.6-9 REVISION 14

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.4 REQUIREMENTS (continued) A COT is performed on each required channel to ensure the entire channel will perform the intended Function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. For MODES 1, 2, 3, and 4, this test verifies the capability of the instrumentation to provide the containment purge and exhaust system isolation. During CORE ALTERATIONS and movement of recently irradiated fuel in containment, this test verifies the capability of the required channels to generate the signals required for input to the control room alarm. There is a plant specific program which verifies that the instrument channel functions as required by verifying the as-left and as-found setting are consistent with those established by the setpoint methodology.

SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.

For slave relays and associated auxiliary relays in the CVI actuation system circuit that are Potter and Brumfield (P&B) type Motor Driven Relays (MDR), the SLAVE RELAY TEST is performed on an 18-month frequency. This test frequency is based on relay reliability assessments presented in WCAP-13878, Reliability Assessment of Potter and Brumfield MDR Series Relays. The reliability assessments are relay specific and apply only to Potter and Brumfield MDR series relays.

Quarterly testing of the slave relays associated with non-P&B MDR auxiliary relays will be administratively controlled until an alternate method of testing the auxiliary relays is developed or until they are replaced by P&B MDR series relays.

SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADOT. This test is a check of the Manual Actuation Functions. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test (continued)

Vogtle Units 1 and 2 B 3.3.6-11 REVISION 20

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES REFERENCES 1. 10 CFR 100.1150.67.

Vogtle Units 1 and 2 B 3.3.6-14 REVISION 14

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity Complete Replacement of the Existing 3.4.16 Bases BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2).

APPLICABLE The LCO limits on the specific activity of the reactor coolant ensure that SAFETY the resulting offsite and control room doses meet the appropriate SRP ANALYSES acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 Ci/gm DOSE EQUIVALENT I-131 from LCO 3.7.18, "Secondary Specific Activity."

The analyses for the SLB and SGTR accidents establish the acceptance limits for RCS specific activity. Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 Ci/gm DOSE EQUIVALENT I-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60.0 Ci/gm DOSE EQUIVALENT I-131 due to an iodine spike caused by a reactor or an RCS transient prior Vogtle Units 1 and 2 B 3.4.16-1 Rev.

BASES APPLICABLE SAFETY ANALYSES (continued) to the accident. In both cases, the noble gas specific activity is assumed to be 280 Ci/gm DOSE EQUIVALENT XE-133.

The SGTR analysis also assumes a loss of offsite power at the same time as the reactor trip. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature T signal.

The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves.

The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the Residual Heat Removal (RHR) system is placed in service.

The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. Reactor trip occurs after the generation of an SI signal on low steam line pressure.

The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the RHR system is placed in service.

Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 60.0 Ci/gm for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The iodine specific activity in the reactor coolant is limited to 1.0 Ci/gm DOSE EQUIVALENT I-131, and the noble gas specific activity in the reactor coolant is limited to 280 Ci/gm DOSE EQUIVALENT XE-133.

The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

The SLB and SGTR accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).

BASES APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 is necessary to Vogtle Units 1 and 2 B 3.4.16-2 Rev.

limit the potential consequences of a SGTR to within the SRP acceptance criteria (Ref. 2).

In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT I-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the specific activity is < 60.0 Ci/gm. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide a trend.

The DOSE EQUIVALENT I-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Actions A.1 and A.2 while the DOSE EQUIVALENT I-131 LCO limit is not met.

This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

Vogtle Units 1 and 2 B 3.4.16-3 Rev.

BASES ACTIONS (continued)

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODES(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

C.1 and C.2 If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT I-131 is > 60.0 Ci/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity. A Frequency Note limits this periodic measurement to MODE 1 when reactor power produces the fission products.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The 7 day Frequency considers the low probability of a gross fuel failure during this time.

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the noble gas specific activity levels peak during this time; samples at other times would provide inaccurate results.

Vogtle Units 1 and 2 B 3.4.16-4 Rev.

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days. A Frequency Note limits this periodic measurement to MODE 1 when reactor power produces the fission products. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

REFERENCES

1. 10 CFR 50.67.

2. Standard Review Plan (SRP) Section 15.0.1 "Radiological Consequence Analyses Using Alternative Source Terms."

3. FSAR, Section 15.1.5.

4. FSAR, Section 15.6.3.

Vogtle Units 1 and 2 B 3.4.16-5 Rev.

SG Tube Integrity B 3.4.17 BASES (continued)

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting SAFETY ANALYSES design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 50.67 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. Portions of the tube below 15.2 inches below the top of the (continued)

Vogtle Units 1 and 2 B 3.4.17-2 REVISION 28

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE SR 3.4.17.2 (continued)

REQUIREMENTS criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES 1. NEI 97-06, Steam Generator Program Guidelines.

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 10050.67.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976.

6. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines.

7. License Amendment Nos. 167 and 149, "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments Regarding Revision to Technical Specifications 5.5.9, "Steam Generator (SG) Program," and 5.6.10, "Steam Generator Tube Inspection Report," (TAC Nos. ME8313 and ME8314),

September 10, 2012.

Vogtle Units 1 and 2 B 3.4.17-7 REVISION 28

Recirculation Fluid pH Control System B 3.5.6 BASES BACKGROUND A long term recirculation solution pH of 7.0 to 10.5 also serves to (continued) minimize the hydrogen produced by the corrosion of galvanized surfaces and zinc-based paints.

In addition, the determination of this pH range also considered the environmental qualification of equipment in containment that may be subjected to the containment spray.

In order to achieve the desired pH range of 7.0 to 10.5 in the post-LOCA recirculation solution, a total of between 11,484 pounds (220 ft3) and 14,612 pounds (260 ft3) of TSP is required. The three TSP storage baskets are designed and located to permit the TSP to be dissolved into the containment recirculation sump solution as the post-LOCA water level rises. The stainless steel mesh screen storage baskets are located in the containment sump area anchored to the filler slab at elevation 171-ft 9-in. The post-LOCA ECCS recirculation and containment spray provide mixing to achieve a uniform solution pH.

TSP, because of its stability when exposed to radiation and elevated temperature and its nontoxic nature, is the preferred buffer material.

The dodecahydrate form of TSP is used because of the high humidity in the containment during normal operation. Since the TSP is hydrated, it will not absorb large amounts of water from the humid atmosphere and will be less susceptible to physical and chemical change than the anhydrous form of TSP.

APPLICABLE Following the assumed release of radioactive material from a SAFETY ANALYSES DBA to the containment atmosphere, the containment is assumed to leak at its design value. The LOCA radiological dose analysis assumes the amount of radioactive material available for release to the outside atmosphere is reduced by the operation of the containment spray system. The analysis also assumes the long term pH control of the recirculation fluid retains the dissolved iodine in solution which prevents the iodine from becoming available for release to the atmosphere (Ref. 2). The radiological consequences of a LOCA may be increased if the long term pH of the recirculation solution is not adjusted to 7.0 or greater. Therefore, long term pH control of the post-LOCA recirculation fluid helps ensure the offsite and control room thyroid doses are within the limits of 10 CFR 100 50.67 and (continued)

Vogtle Units 1 and 2 B 3.5.6-2 REVISION 44

Containment Air Locks 3.6.2 BASES APPLICABLE containment was designed with an allowable leakage rate of SAFETY ANALYSES 0.2% of containment air weight per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued) and 0.1% per day thereafter (Ref. 2). This leakage rate is defined as La = 0.2% of containment air weight per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure Pa = 37 psig following a DBA. This allowable leakage rate (0.2%) forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.

The containment air locks satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO Each containment air lock forms part of the containment pressure boundary. As part of containment, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.

Closure of a single door in each air lock is sufficient to provide a leaktight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. The pressure and temperature limitations of MODES 5 and 6 reduce the probability and consequences of the events events considered for MODES 1, 2, 3, and 4. Therefore, the containment air locks are not required in MODE 5 to prevent leakage of radioactive material from containment. In MODE 6, the limiting location for dose consequences resulting from a fuel handling accident is in the fuel building and not containment. As a result, no credit is taken for the closure of the airlocks.the requirements for (continued)

Vogtle Units 1 and 2 B 3.6.2-2 Rev. 1-10/01

Containment Air Locks 3.6.2 BASES APPLICABILITY the containment air locks are based on a fuel handling accident inside (continued) containment. The requirements for the containment air locks during MODE 6 are addressed in LCO 3.9.4, "Containment Penetrations."

ACTIONS The ACTIONS are modified by a Note that allows entry and exit to perform repairs on the affected air lock component. If the outer door Is inoperable, then it may be easily accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the containment boundary is not intact (during access through the outer door). The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the containment during the short time in which the OPERABLE door is expected to be open. After each entry and exit, the OPERABLE door must be immediately closed.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each air lock.

In the event the air lock leakage results in exceeding the overall containment leakage rate, Note 3 directs entry into the applicable Conditions and Required Actions of LCO 3.6.1, "Containment."

A.1, A.2, and A.3 With one air lock door in one or more containment air locks inoperable, the OPERABLE door must be verified closed (Required Action A.1) in each affected containment air lock. This ensures that a leak tight containment barrier is maintained by the use of an OPERABLE air lock door. This action must be completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires containment be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(continued)

Vogtle Units 1 and 2 B 3.6.2-3 Rev. 1-8/98

MSIVs B 3.7.2 BASES APPLICABLE limiting with respect to the steam releases used in meeting SAFETY ANALYSES equipment qualification criteria. The failure of an MSIV has no (continued) effect on the results of these events.

c. A break downstream of the MSIVs will be isolated by the closure of the MSIVs. This is not a limiting scenario with respect to doses or with respect to the core response analyses.

d. For a steam generator tube rupture, closure of the MSIVs in the faulted loop isolates the ruptured steam generator from the intact steam generators to minimize radiological releases.

The MSIVs satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO This LCO requires that two MSIV systems in each steam line be OPERABLE. The MSIV systems are considered OPERABLE when the isolation times are within limits, and they close on an isolation actuation signal. An OPERABLE MSIV system may consist of an OPERABLE MSIV and inoperable associated bypass valve provided the inoperable bypass valve is maintained closed.

This LCO provides assurance that the MSIV systems will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the 10 CFR 100 50.67 (Ref. 6) limits or the NRC staff approved licensing basis.

APPLICABILITY The MSIV systems must be OPERABLE in MODE 1, and in MODES 2 and 3 except when one MSIV system in each steam line is closed, when there is significant mass and energy in the RCS and steam generators. When the MSIV systems are closed, they are already performing the safety function.

In MODE 4, normally most of the MSIV systems are closed, and the steam generator energy is low.

(continued)

Vogtle Units 1 and 2 B 3.7.2-3 Rev. 1-10/01

MSIVs B 3.7.2 BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV system is within the limit given in Reference 9 on an actual or simulated actuation signal and is within that assumed in the accident and containment analyses. This SR also verifies the valve closure time is in accordance with the INSERVICE TESTING PROGRAM. This SR is normally performed upon returning the unit to operation following a refueling outage.

The Frequency is in accordance with the INSERVICE TESTING PROGRAM. Operating experience has shown that these components usually pass the Surveillance when performed in accordance with the INSERVICE TESTING PROGRAM. Therefore, the Frequency is acceptable from a reliability standpoint.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. If desired, this allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

REFERENCES 1. FSAR, Section 10.3.

2. FSAR, Section 6.2.

3. FSAR, Subsection 15.1.5.

4. FSAR, Subsection 15.4.9.

5. FSAR, Subsection 15.2.8.

6. 10 CFR 100.1150.67.

7. ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code).

8. Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Implementation of Topical Report Nuclear Energy Institute NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A (CAC Nos.

ME9555 and ME9556).

9. TRM, Section 13.7.6.

Vogtle Units 1 and 2 B 3.7.2-7 REVISION 64

ARVs B 3.7.4 BASES APPLICABLE event and other accident analyses. After primary to SAFETY ANALYSES secondary break flow termination, it is assumed that one ARV (continued) on an intact SG is used to cool the RCS down to 350°F, at the maximum allowable cooldown rate of 100°F/hour.

The offsite radiological dose analyses show that the failure open of the ARV on the ruptured SG represents the limiting single failure. The resulting offsite radiological doses at the exclusion area boundary, low population zone, and control room are well within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10 CFR 10050.67. A detailed description of the SGTR analyses can be found in WCAP-11731 and associated supplements (Ref. 3).

The ARVs are equipped with manual block valves in the event an ARV spuriously fails open or fails to close during use.

The ARVs satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO Three ARV lines are required to be OPERABLE. One ARV line is required from each of three steam generators to ensure that at least one ARV line is available to conduct a unit cooldown following an SGTR, in which one steam generator becomes unavailable, accompanied by a single, active failure of a second ARV line on an unaffected steam generator. A block valve for each required ARV must be OPERABLE to isolate a failed open ARV line.

Failure to meet the LCO can result in the inability to cool the unit to RHR entry conditions following an SGTR event in which the condenser is unavailable for use with the Steam Dump System.

An ARV is considered OPERABLE when it is capable of providing controlled relief of the main steam flow and capable of fully opening and closing on demand. Additionally, it is required that at least two of the three OPERABLE ARVs maintain the capability for local manual actuation via their associated handpumps.

APPLICABILITY In MODES 1, 2, and 3, the ARVs are required to be OPERABLE.

In MODE 4, the pressure and temperature limitations are such that the probability of an SGTR event requiring ARV operation is low. In addition, the RHR system is available (continued)

Vogtle Units 1 and 2 B 3.7.4-3 Rev. 2-10/01

PPAFES B 3.7.13 BASES BACKGROUND moisture removal. The primary purpose of the heaters is to (continued) maintain the relative humidity at an acceptable level; however, the VEGP dose analysis assumes no heater operation and an iodine removal efficiency consistent with the iodine removal efficiency in Regulatory Guide 1.52 (Ref. 4) for systems designed to operate inside primary containment (i.e., no humidity control). Therefore, the heaters are not required for PPAFES OPERABILITY.

APPLICABLE The PPAFES design basis is established by the large break SAFETY ANALYSES loss of coolant accident (LOCA). The system evaluation assumes 2 gpm continuous leakage and a 50 gpm leak for 30 minutes due to a passive failure during a Design Basis Accident (DBA). The system restricts the radioactive release to within the 10 CFR 100 50.67 (Ref. 4) limits, or the NRC staff approved licensing basis (e.g., a specified fraction of 10 CFR 100 50.67 limits). The analysis of the effects and consequences of a large break LOCA are presented in Reference 3.

The PPAFES satisfies Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO Two independent and redundant trains of the PPAFES are required to be OPERABLE to ensure that at least one train is available, assuming there is a single failure disabling the other train coincident with a loss of offsite power.

The PPAFES is considered OPERABLE when the individual components necessary to control radioactive releases are OPERABLE in both trains. A PPAFES train is considered OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration functions; and

c. Demister, ductwork, valves, and dampers are OPERABLE and air circulation can be maintained.

The LCO is modified by a Note allowing the PPAFES boundary to be opened intermittently under administrative controls without requiring entry into the Condition for an inoperable pressure boundary. For (continued)

Vogtle Units 1 and 2 B 3.7.13-2 Rev. 2-10/01

PPAFES B 3.7.13 BASES LCO entry and exit through doors, the administrative control of the opening (continued) is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for PPAFES isolation is indicated.

APPLICABILITY In MODES 1, 2, 3, and 4, the PPAFES is required to be OPERABLE, consistent with the OPERABILITY requirements of the ECCS.

In MODE 5 or 6, the PPAFES is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

ACTIONS A.1 With one PPAFES train inoperable, the action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the PPAFES function. The 7 day Completion Time is appropriate because the risk contribution of the PPAFES is less than that of the ECCS (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this period, and the remaining train providing the required capability.

B.1 If the PPAFES boundary is inoperable, the PPAFES trains cannot perform their intended function. Actions must be taken to restore an OPERABLE PPAFES boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the period that the PPAFES boundary is inoperable, appropriate compensatory measures (consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR 10050.67) will be utilized to ensure the necessary physical security and to minimize the release of radioactive material to the atmosphere outside the building. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24-hour Completion Time is reasonable based on the low (continued)

Vogtle Units 1 and 2 B 3.7.13-3 Rev. 1-5/01

PPAFES B 3.7.13 BASES (continued)

REFERENCES 1. FSAR, Subsection 6.5.1.

2. FSAR, Subsection 9.4.3.

3. FSAR, Subsection 15.6.5.

4. 10 CFR 10050.67.

5. WCAP-16294-NP-A, Rev. 1, Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs, June 2010.

Vogtle Units 1 and 2 B 3.7.13-7 REVISION 72

Fuel Storage Pool Water Level B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Subsection 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Subsection 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Subsection 15.7.4 (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets SAFETY ANALYSES the assumptions of the fuel handling accident described in Regulatory Guide 1.195 183 (Ref. 4). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) limitsless than the acceptance criteria specified in Regulatory Guide 1.183, Revision 0 (Reference 4).

According to Reference 4,If there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident, then. With 23 ft of water, the assumptions ofdecontamination factors specified in Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle.

To offset this small non-conservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop. The analyses also assume a limited number of fuel rods are damaged in a second fuel bundle.

The fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

(continued)

Vogtle Units 1 and 2 B 3.7.15-1 REVISION 31

Fuel Storage Pool Water Level B 3.7.15 BASES SURVEILLANCE SR 3.7.15.1 (continued)

REQUIREMENTS water level in the fuel storage pool must be checked periodically.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

REFERENCES 1. FSAR, Subsection 9.1.2.

2. FSAR, Subsection 9.1.3.

3. FSAR, Subsection 15.7.4.

4. Regulatory Guide 1.195183 Revision 0, Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors, JulyMay 20003.

5. 10 CFR 100.11.

Vogtle Units 1 and 2 B 3.7.15-3 REVISION 31

Secondary Specific Activity B 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions.

During transients, I-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 1 gpm tube leak (LCO 3.4.13, "RCS Operational LEAKAGE") of primary coolant at the limit of 1.0 Ci/gm (LCO 3.4.16, "RCS Specific Activity"). The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and the reactor coolant LEAKAGE. Most of the iodine isotopes have short half lives, (i.e., < 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ). I-131, with a half life of 8.04 days, concentrates faster than it decays, but does not reach equilibrium because of blowdown and other losses.

With the specified activity limit, the resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the exclusion area boundary (EAB) would be about 0.58 rem if the main steam safety valves (MSSVs) open for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a trip from full power.

Operating a unit at the allowable limits could result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months EAB exposure of a small fraction of the 10 CFR 100 50.67 (Ref. 1) limits.

(continued)

Vogtle Units 1 and 2 B 3.7.16-1 Revision No. 0

Secondary Specific Activity B 3.7.16 BASES (continued)

REFERENCES 1. 10 CFR 100.1150.67.

2. FSAR, Chapter 15.

Vogtle Units 1 and 2 B 3.7.16-4 Revision No. 0

Boron Concentration B 3.9.1 BASES APPLICABLE The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36 SAFETY ANALYSES (c)(2)(ii).

(continued)

LCO The LCO requires that a minimum boron concentration be maintained in all filled portions of the RCS, the refueling canal, and the refueling cavity while in MODE 6. The boron concentration limit specified in the COLR ensures that a core keff of 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6.

APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a keff 0.95. In MODES 1 and 2, LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits," ensure an adequate amount of negative reactivity is available to shut down the reactor. In MODES 3, 4, and 5, LCO 3.1.1, "SHUTDOWN MARGIN" ensures an adequate amount of negative reactivity is available to shut down the reactor.

The Applicability is modified by a Note. The Note states that the limits on boron concentration are only applicable to the refueling canal and the refueling cavity when those volumes are connected to the Reactor Coolant System. When the refueling canal and the refueling cavity are isolated from the RCS, no potential path for boron dilution exists.

ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the filled portions of the RCS, the refueling canal, or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately.

(continued)

Vogtle Units 1 and 2 B 3.9.1-3 REVISION 40

Boron Concentration B 3.9.1 BASES ACTIONS A.1 and A.2 (continued)

Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position or normal cooldown of the coolant volume for the purpose of system temperature control.

A.32 In addition to immediately suspending CORE ALTERATIONS or positive reactivity additions, boration to restore the concentration must be initiated immediately.

There are no safety analysis assumptions of boration flow rate and concentration that must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in all filled portions of the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis. Prior to re-connecting portions of the refueling canal or the refueling cavity to the RCS, this SR must be met per SR 3.0.4. If any dilution has occurred while the cavity or canal were disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Subsection 15.4.6.

Vogtle Units 1 and 2 B 3.9.1-4 REVISION 40

Unborated Water Source Isolation Valves B 3.9.2 BASES LCO administrative control provided the reactor coolant system boron (continued) concentration is within the limit specified in the COLR and the high flux at shutdown alarm is OPERABLE. The high flux at shutdown alarm is not normally required OPERABLE in MODE 6, however for the purpose of meeting the requirement stated in this Note, the high flux at shutdown alarm is considered OPERABLE if the applicable surveillance requirements of LCO 3.3.8, High Flux at Shutdown Alarm and LCO 3.9.3, Nuclear Instrumentation are met.

APPLICABILITY In MODE 6, this LCO is applicable to prevent an inadvertent boron dilution event by ensuring isolation of all sources of unborated water to the RCS.

For all other MODES, the boron dilution accident was analyzed and was found to be capable of being mitigated.

ACTIONS The ACTIONS do not apply to valves in the flow path from the RMWST, through the chemical mixing tank, to the suction of the charging pumps, when opened under administrative control in accordance with the Note in the LCO. The ACTIONS table has been modified by a Note that allows separate Condition entry for each unborated water source isolation valve.

A.1 Continuation of CORE ALTERATIONS is contingent upon maintaining the unit in compliance with this LCO. With any valve used to isolate unborated water sources not secured in the closed position, all operations involving CORE ALTERATIONS must be suspended immediately. The Completion Time of "immediately" for performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position.

Condition A has been modified by a Note to require that Required Action A.3 be completed whenever Condition A is entered.

(continued)

Vogtle Units 1 and 2 B 3.9.2-3 Revision No. 0

Unborated Water Source Isolation Valves B 3.9.2 BASES ACTIONS A.21 (continued)

Preventing inadvertent dilution of the reactor coolant boron concentration is dependent on maintaining the unborated water isolation valve(s) secured closed. Securing the valve(s) in the closed position ensures that the valve(s) cannot be inadvertently opened.

The Completion Time of "immediately" requires an operator to initiate actions to close an open valve and secure the isolation valve in the closed position immediately. Once actions are initiated, they must be continued until the valves are secured in the closed position.

A.32 Due to the potential of having diluted the boron concentration of the reactor coolant, SR 3.9.1.1 (verification of boron concentration) must be performed whenever Condition A is entered to demonstrate that the required boron concentration exists. The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to obtain and analyze a reactor coolant sample for boron concentration.

SURVEILLANCE SR 3.9.2.1 REQUIREMENTS These valve(s) are to be secured closed to isolate possible dilution paths. The likelihood of a significant reduction in the boron concentration during MODE 6 operations is remote due to the large mass of borated water in the refueling cavity and the fact that all unborated water sources are isolated, precluding a dilution. The boron concentration is checked every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during MODE 6 under SR 3.9.1.1. This Surveillance demonstrates that the valves are closed through a system walkdown. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. FSAR, Subsection 15.4.6.

2. NUREG-0800, Section 15.4.6.

Vogtle Units 1 and 2 B 3.9.2-4 REVISION 14

Nuclear Instrumentation B 3.9.3 BASES LCO source of power, provided the detector for the opposite source range (continued) neutron flux monitor is powered from its normal source.

APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, the operability requirements for the installed source range detectors and circuitry are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation."

ACTIONS A.1 and A.2 With only one source range neutron flux monitor OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position or normal cooldown of the coolant volume for the purpose of system temperature control. Suspending the movement of fuel, sources, and reactivity control components ensures that positive reactivity is not inadvertently added to the reactor core while the source range neutron flux monitor is inoperable. Required Action A.1 is modified by a Note that states that fuel assemblies, sources, and reactivity control components may be moved if necessary to facilitate repair or replacement of the inoperable source range neutron flux monitor. It may be necessary to move these items away from the locations in the core close to the source range neutron flux monitor to minimize personnel radiation dose during troubleshooting or repair. The Note also permits completion of movement of a component to a safe position, should the source range neutron flux monitor be discovered inoperable during component movement.

B.1 Condition B is modified by a Note to clarify the requirement that entry into or continued operation in accordance with Condition A is required for any entry into Condition B. The Note reinforces conventions of LCO applicability as stated in LCO 3.0.2 and as reflected in examples in 1.3, Completion Times.

(continued)

Vogtle Units 1 and 2 B 3.9.3-2 Rev. 1-4/09

Nuclear Instrumentation B 3.9.3 With no source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately.

Once initiated, actions shall be continued until a source range neutron flux monitor is restored to OPERABLE status.

B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS (except as allowed by the Note to Required Action A.1) and positive reactivity additions are not to be (continued)

Vogtle Units 1 and 2 B 3.9.3-3 Rev. 1-4/09

Containment Penetrations B 3.9.4 B 3.9 REFUELING OPERATIONS B 3.9.4 Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY."

Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50, Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements limits of 10 CFR 100Regulatory Guide 1.183 (Reference 3). Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. If closed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced. Alternatively, the equipment hatch can be open provided it can be installed with a minimum of four bolts holding it in place.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is (continued)

Vogtle Units 1 and 2 B 3.9.4-1 Rev. 1-9/00

Containment Penetrations B 3.9.4 BASES BACKGROUND required. During periods of unit shutdown when containment (continued) closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment, the door interlock mechanism may remain disabled, but one air lock door must always must be isolable by at least one air lock door with a designated individual available to close the air lock door, or at least one air lock door must be closed.

The emergency air lock will not normally be open during core alterations or fuel movement inside containment. Therefore, in the event the emergency air lock is open at the same time the personnel air lock is open, a separate individual shall be responsible for closing the emergency air lock (within 15 minutes) in addition to the individual designated to close the personnel air lock.

The requirements for containment penetration closure are sufficient to ensure fission product radiactivityradioactivity release from containment due to a fuel handling accident involving handling of recently irradiated fuel during refueling is maintained to within the acceptance criteria of Standard Review Plan Section 15.7.4 and General Design Criteria 19.

The Containment Ventilation System consists of two 24 inch penetrations for purge and exhaust of the containment atmosphere. Each main or shutdown purge and exhaust system contains one motor operated 24 inch valve inside containment and one motor operated 24 inch valve outside containment (HV-2626A, HV-2627A, HV-2628A, and HV-2629A). A second 14 inch mini-purge and exhaust system shares each 24 inch penetration and consists of one 14 inch pneumatically operated valve inside containment and one outside of containment (HV-2626B, HV-2627B, HV-2628B, and HV-2629B). A 14 inch mini-purge line is connected to each 24 inch line between the 24 inch isolation valve and the penetration both inside and outside containment.

In MODES 1, 2, 3 and 4 the 24 inch main or shutdown purge and exhaust valves are secured in the closed position. The 14 inch mini-purge and exhaust valves may be opened in these MODES in accordance with LCO 3.6.3, Containment Isolation Valves, and are automatically closed by a Containment Ventilation Isolation signal. The instrumentation that provides the automatic isolation function for these valves is listed in LCO 3.3.6, Containment Ventilation Isolation Instrumentation.

(continued)

Vogtle Units 1 and 2 B 3.9.4-3 Rev. 1-3/99

Containment Penetrations B 3.9.4 BASES BACKGROUND In MODE 6, the 24 inch main or shutdown purge and exhaust (continued) valves are used to exchange large volumes of containment air to support refueling operations or other maintenance activities. During CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment any open 24 inch valves are capable of being closed (LCO 3.3.6). The 14 inch mini-purge and exhaust valves, though typically not opened during CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment, if opened are also capable of being closed (LCO 3.3.6).

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by a closed automatic isolation valve, a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods allowed under the provisions of 10 CFR 50.59 may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment (Ref. 1).

APPLICABLE During CORE ALTERATIONS or movement of recently irradiated fuel SAFETY ANALYSES assemblies within containment, the most severe radiological consequences result from a fuel handling accident involving recently irradiated fuel. The fuel handling accident is a postulated event that involves damage to recently irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in Reference 2, include dropping a single irradiated fuel assembly onto another irradiated fuel assembly.

To support the plant configuration of both air lock doors open (personnel and/or emergency air locks), and to further minimize an unmonitored, untreated release, the designated individual for closure of the air lock will have the air lock closed within 15 minutes of the fuel handling accident. The 15 minute duration was chosen as the limit for the response capability for the person who is designated for closing the air lock door. The NRC (continued)

Vogtle Units 1 and 2 B 3.9.4-4 REVISION 31

Containment Penetrations B 3.9.4 BASES APPLICABLE acceptance of this specification was based on doses for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months SAFETY ANALYSES release as well as a licensee commitment for a person (continued) designated to close the door quickly.

The requirements of LCO 3.9.7, "Refueling Cavity Water Level," and thein conjunction with minimum decay time of 90 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months prior to CORE ALTERATIONS irradiated fuel movement with containment closure capability or a minimum decay time of 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months without containment closure capability ensures that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100less than the acceptance criteria specified in Regulatory Guide 1.183, Revision 0 (Reference 3). The acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values as specified in Regulatory Guide 1.195 (Ref. 3). The radiological consequences of a fuel handling accident in containment have been evaluated assuming that the containment is open to the outside atmosphereis non-limiting as compared to a fuel handling accident in the fuel building. All airborne activity reaching the containment atmosphere is assumed to be exhausted to the environment within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the accident. The calculated offsite and control room operator doses are within the acceptance criteria of Regulatory Guide 1.195 and GDC 19. Therefore, although the containment penetrations do not satisfy any of the 10 CFR 50.36 (c)(2)(ii) criteria, LCO 3.9.4 provides containment closure capability to minimize potential offsite doses.

LCO This LCO limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires the equipment hatch, the air locks, and any penetration providing direct access to the outside atmosphere to be closed or capable of being closed. Personnel air lock closure capability is provided by the availability of at least one door and a designated individual to close it. Emergency air lock closure capability is provided by the availability of at least one door and a designated individual to close it. Equipment hatch closure capability is provided by a designated trained hatch closure crew and the necessary equipment. For the OPERABLE containment ventilation penetrations, this LCO ensures that each penetration is isolable by the Containment Ventilation Isolation valves. The OPERABILITY requirements for LCO 3.3.6, Containment Ventilation Isolation Instrumentation ensure that radiation monitor inputs to the control room alarm exist so that operators can take timely (continued)

Vogtle Units 1 and 2 B 3.9.4-5 REVISION 31

Containment Penetrations B 3.9.4 BASES LCO action to close containment penetrations to minimize potential offsite (continued) doses. The LCO requirements for penetration closure may also be met by the automatic isolation capability of the CVI system.

Temporary non-1E power may be supplied to the air operated and/or solenoid operated CVI valves. The temporary non-1E power must be connected in such a way that it cannot affect the capability of the valves to close either automatically or manually from the control room handswitch.

The LCO is modified by a Note allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls.

Administrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident.

Item b of this LCO includes requirements for both the emergency air lock and the personnel air lock. The personnel and emergency air locks are required by Item b of this LCO to be isolable by at least one air lock door in each air lock. Both containment personnel and emergency air lock doors may be open during movement of recently irradiated fuel in the containment and during CORE ALTERATIONS provided at least one air lock door is isolable in each air lock. An air lock is isolable when the following criteria are satisfied:

1. one air lock door is OPERABLE,

2. at least 23 feet of water shall be maintained over the top of the reactor vessel flange in accordance with Specification 3.9.7,

3. a designated individual is available to close the door.

OPERABILITY of a containment air lock door requires that the door seal protectors are easily removed, that no cables or hoses are being run through the air lock, and that the air lock door is capable of being quickly closed.

The equipment hatch is considered isolable when the following criteria are satisfied:

1. the necessary equipment required to close the hatch is available.

(continued)

Vogtle Units 1 and 2 B 3.9.4-7 REVISION 41

Containment Penetrations B 3.9.4 BASES LCO 2. at least 23 feet of water is maintained over the top of the reactor (continued) vessel flange in accordance with Specification 3.9.7,

3. a designated trained hatch closure crew is available.

Similar to the air locks, the equipment hatch opening must be capable of being cleared of any obstruction so that closure can be achieved as soon as possible.

APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment because this is when there is a potential for a the limiting fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1, "Containment." In MODES 5 and 6, when CORE ALTERATIONS or movement of recently irradiated fuel assemblies within containment are is not being conducted, the potential for a fuel handling accident does not exist. Additionally, due to radioactive decay, a fuel handling accident involving handing recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months ) will result in does that are well within the guideline values specified in 10 CFR 100 even without containment closure capability. Therefore, under these conditions no requirements are placed on containment penetration status.

ACTIONS A.1 and A.2 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of recently irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position.

The Surveillance on the required open containment ventilation isolation valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will demonstrate that each required (continued)

Vogtle Units 1 and 2 B 3.9.4-8 REVISION 41

Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR 3.9.4.2 REQUIREMENTS (continued) This Surveillance demonstrates that each containment ventilation isolation valve in each open containment ventilation penetration actuates to its isolation position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note stating that this surveillance is not required to be met for valves in isolated penetrations. LCO 3.9.4.c.1 provides the option to close penetrations in lieu of requiring automatic actuation capability.

SR 3.9.4.3 The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

The 7 day frequency is adequate considering that the hardware, tools, and equipment are dedicated to the equipment hatch and not used for any other functions.

The SR is modified by a Note which only requires that the surveillance be met for an open equipment hatch. If the equipment hatch is installed in its opening, the availability of the means to install the hatch is not required.

REFERENCES 1. GPU Nuclear Safety Evaluation SE-0002000-001, Rev. 0, May 20, 1988.

2. FSAR, Subsection 15.7.4.

3. Regulatory Guide 1.195, May 2003Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents At Nuclear Reactors, July 2000.

Vogtle Units 1 and 2 B 3.9.4-10 REVISION 40

Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies or performance of CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to < 25% of 10 CFR 100 limits, as provided by the guidance of Reference 3less than the acceptance criteria in Reference 1.

APPLICABLE During CORE ALTERATIONS and movement of irradiated fuel SAFETY ANALYSES assemblies, the water level in the refueling canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment, as postulated by Regulatory Guide 1.195 183 (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water.

The fuel handling accident analysis inside containment is non-limiting compared to a fuel handling accident in the fuel building as described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 90 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained less than the acceptance criteria specified in Reference 1within allowable limits (Refs. 3 and 4).

(continued)

Vogtle Units 1 and 2 B 3.9.7-1 REVISION 31

Refueling Cavity Water Level B 3.9.7 BASES APPLICABLE Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36 SAFETY ANALYSES (c)(2)(ii).

(continued)

LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits, as provided by the acceptance criteria in Reference 1.guidance of Reference 3.

APPLICABILITY LCO 3.9.7 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and when moving irradiated fuel assemblies within containment. Unlatching and latching of control rod drive shafts includes drag testing of the associated rod cluster control assembly. The LCO ensures a sufficient level of water is present in the reactor cavity to minimize the radiological consequences of a fuel handling accident in containment. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.15, "Fuel Storage Pool Water Level."

ACTIONS A.1 and A.2 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.

(continued)

Vogtle Units 1 and 2 B 3.9.7-2 Rev. 1-10/01

Refueling Cavity Water Level B 3.9.7 BASES (continued)

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. Regulatory Guide 1.195, May 2003183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

2. FSAR, Subsection 15.7.4.

3. 10 CFR 100.11

4. Malinowski, D. D., Bell, M. J., Duhn, E., and Locante, J.,

WCAP-7828, Radiological Consequences of a Fuel Handling Accident, December 1971.

Vogtle Units 1 and 2 B 3.9.7-3 REVISION 31 to Enclosure Regulatory Guide 1.183 Conformance Tables Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 4 Regulatory Guide 1.183 Conformance Tables A4-1 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 1.1.1 The proposed uses of an AST and the associated Conforms- Adequate proposed facility modifications and changes to safety margins are procedures should be evaluated to determine whether maintained, as the proposed changes are consistent with the principle discussed in the No that sufficient safety margins are maintained, including a Significant Hazards margin to account for analysis uncertainties. The safety Consideration. Future margins are products of specific values and limits changes will be contained in the technical specifications (which cannot evaluated under the be changed without NRC approval) and other values, provisions of 10 CFR such as assumed accident or transient initial conditions 50.59.

or assumed safety system response times. Changes, or the net effects of multiple changes, that result in a reduction in safety margins may require prior NRC approval. Once the initial AST implementation has been approved by the staff and has become part of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in assessing safety margins related to subsequent facility modifications and changes to procedures.

1.1.2 The proposed uses of an AST and the associated Conforms - There are proposed facility modifications and changes to no facility modifications procedures should be evaluated to determine whether being proposed to the proposed changes are consistent with the principle implement AST, and that adequate defense in depth is maintained to compliance with the compensate for uncertainties in accident progression GDCs are maintained.

and analysis data. Consistency with the defense-in- No new reliance is depth philosophy is maintained if system redundancy, placed on independence, and diversity are preserved compensatory commensurate with the expected frequency, programmatic actions consequences of challenges to the system, and (including manual uncertainties. In all cases, compliance with the General operator actions) to Design Criteria in Appendix A to 10 CFR Part 50 is maintain adequate essential. Modifications proposed for the facility defense-in-depth.

generally should not create a need for compensatory programmatic activities, such as reliance on manual operator actions.

1.1.2 Proposed modifications that seek to downgrade or Not Applicable - There remove required engineered safeguards equipment are no modifications should be evaluated to be sure that the modification being proposed with does not invalidate assumptions made in facility PRAs this License and does not adversely impact the facility's severe Amendment Request.

accident management program.

1.1.3 The design basis accident source term is a fundamental Conforms - See RG assumption upon which a significant portion of the Section 1.3 facility design is based. Additionally, many aspects of discussions.

facility operation derive from the design analyses that A4-2 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis incorporated the earlier accident source term. Although a complete re-assessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses would generally not be necessary. Regulatory Position 1.3 of this guide provides guidance on which analyses need updating as part of the AST implementation submittal and which may need updating in the future as additional modifications are performed.

1.1.3 This approach would create two tiers of analyses, those Conforms - This is a based on the previous source term and those based on full scope AST an AST. The radiological acceptance criteria would also implementation for the be different with some analyses based on whole body radiological dose and thyroid criteria and some based on TEDE criteria. consequences of the Full implementation of the AST revises the plant VEGP Design Basis licensing basis to specify the AST in place of the Accidents.

previous accident source term and establishes the TEDE dose as the new acceptance criteria. Selective implementation of the AST also revises the plant licensing basis and may establish the TEDE dose as the new acceptance criteria. Selective implementation differs from full implementation only in the scope of the change. In either case, the facility design bases should clearly indicate that the source term assumptions and radiological criteria in these affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria.

1.1.3 Radiological analyses generally should be based on Conforms- This License assumptions and inputs that are consistent with Amendment Request corresponding data used in other design basis safety includes re-evaluation analyses, radiological and nonradiological, unless these of the radiological data would result in nonconservative results or consequences of the otherwise conflict with the guidance in this guide. most severe DBAs. It relies on assumptions and inputs that do not create a conflict with, or render non-conservative, other design basis safety analyses.

1.1.4 Although the AST provided in this guide was based on a Conforms - No limited spectrum of severe accidents, the particular changes are proposed characteristics have been tailored specifically for DBA in this License A4-3 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis analysis use. The AST is not representative of the wide Amendment Request to spectrum of possible events that make up the planning Emergency basis of emergency preparedness. Therefore, the AST Preparedness is insufficient by itself as a basis for requesting relief requirements.

from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50.

1.2.1 Full implementation is a modification of the facility Conforms - This design basis that addresses all characteristics of the License Amendment AST, that is, composition and magnitude of the Request involves radioactive material, its chemical and physical form, and recalculation of the the timing of its release. Full implementation revises the dose consequences of plant licensing basis to specify the AST in place of the the most severe DBAs.

previous accident source term and establishes the The characteristics of TEDE dose as the new acceptance criteria. This applies the AST methods are not only to the analyses performed in the application addressed in the (which may only include a subset of the plant analyses), recalculations. The but also to all future design basis analyses. At a DBA LOCA has been minimum for full implementations, the DBA LOCA must re-analyzed per be re-analyzed using the guidance in Appendix A of this Appendix A.

guide. Additional guidance on analysis is provided in Regulatory Position 1.3 of this guide. Since the AST and TEDE criteria would become part of the facility design basis, new applications of the AST would not require prior NRC approval unless stipulated by 10 CFR 50.59, Changes, Tests, and Experiments, or unless the new application involved a change to a technical specification. However, a change from an approved AST to a different AST that is not approved for use at that facility would require a license amendment under 10 CFR 50.67.

1.2.2 Selective implementation is a modification of the facility Not Applicable - This design basis that (1) is based on one or more of the License Amendment characteristics of the AST or (2) entails re-evaluation of Request is for full a limited subset of the design basis radiological scope AST analyses. The NRC staff will allow licensees flexibility in implementation for the technically justified selective implementations provided radiological dose a clear, logical, and consistent design basis is consequences of the maintained. An example of an application of selective major VEGP DBA.

implementation would be one in which a licensee desires to use the release timing insights of the AST to increase the required closure time for a containment isolation valve by a small amount. Another example would be a request to remove the charcoal filter media from the spent fuel building ventilation exhaust. For the latter, the licensee may only need to re-analyze DBAs that credited the iodine removal by the charcoal media.

Additional analysis guidance is provided in Regulatory A4-4 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis Position 1.3 of this guide. NRC approval for the AST (and the TEDE dose criterion) will be limited to the particular selective implementation proposed by the licensee. The licensee would be able to make subsequent modifications to the facility and changes to procedures based on the selected AST characteristics incorporated into the design basis under the provisions of 10 CFR 50.59. However, use of other characteristics of an AST or use of TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, would require prior staff approval under 10 CFR 50.67. As an example, a licensee with an implementation involving only timing, such as relaxed closure time on isolation valves, could not use 10 CFR 50.59 as a mechanism to implement a modification involving a reanalysis of the DBA LOCA.

However, this licensee could extend use of the timing characteristic to adjust the closure time on isolation valves not included in the original approval.

1.3.1 There are several regulatory requirements for which Conforms- This full compliance is demonstrated, in part, by the evaluation scope AST License of the radiological consequences of design basis Amendment Request is accidents. These requirements include, but are not salient to: a) Control limited to, the following. Room Habitability Environmental Qualification of Equipment (10 (GDC 19 and NUREG-CFR 50.49) 0737 Item III.D.3.4), b)

Control Room Habitability (GDC 19 of Appendix AST (10 CFR 50.67),

A to 10 CFR Part 50) and c) Facility Siting Emergency Response Facility Habitability (10 CFR 100.11).

(Paragraph IV.E.8 of Appendix E to 10 CFR Part Control Room

50) Habitability and Alternative Source Term (10 CFR 50.67) compliance with the Environmental Reports (10 CFR Part 51) Alternative Source Facility Siting (10 CFR 100.11)5 Term requirements are There may be additional applications of the accident the principal subjects of source term identified in the technical specification this submittal and are bases and in various licensee commitments. These discussed in Sections 3 include, but are not limited to, the following from and 4 of this License Reference 2, NUREG-0737. Amendment Request.

Post-Accident Access Shielding (NUREG-0737, II.B.2) Regarding Emergency Response Facility Post-Accident Sampling Capability (NUREG-Habitability, VEGP will 0737, II.B.3) continue to meet the Accident Monitoring Instrumentation (NUREG-NUREG-0654 Planning 0737, II.F.1)

Standard for Leakage Control (NUREG-0737, III.D.1.1) Emergency Facilities A4-5 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis Emergency Response Facilities (NUREG-0737, and Equipment as III.A.1.2) described in the VEGP Control Room Habitability (NUREG-0737, Emergency Plan.

III.D.3.4) Design Basis dose calculations for non-control room Emergency Response Facilities, such as the Technical Support Center, are not part of the VEGP current licensing basis. The Emergency Response Facilities continue to meet NUREG-0696 habitability requirements.

As stated in Footnote 5 of this RG, the dose guidelines of 10 CFR 100.11 are superseded by 10 CFR 50.67 for licensees that have implemented an AST.

1.3.2 Any implementation of an AST, full or selective, and any Conforms- The License associated facility modification should be supported by Amendment Request evaluations of all significant radiological and for this full scope nonradiological impacts of the proposed actions. This application of the AST evaluation should consider the impact of the proposed evaluated the impact of changes on the facility's compliance with the regulations the proposed change and commitments listed above as well as any other against the Current facility-specific requirements. These impacts may be Licensing Basis, due to (1) the associated facility modifications or (2) the mitigating system differences in the AST characteristics. The scope and design basis extent of the re-evaluation will necessarily be a function requirements, and of the specific proposed facility modification6 and Technical whether a full or selective implementation is being Specifications. No pursued. The NRC staff does not expect a complete facility modifications are recalculation of all facility radiological analyses, but proposed as part of this does expect licensees to evaluate all impacts of the License Amendment proposed changes and to update the affected analyses Request and and the design bases appropriately. An analysis is compliance with considered to be affected if the proposed modification regulations and changes one or more assumptions or inputs used in that commitments are analysis such that the results, or the conclusions drawn maintained.

A4-6 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis on those results, are no longer valid. Generic analyses, such as those performed by owner groups or vendor topical reports, may be used provided the licensee justifies the applicability of the generic conclusions to the specific facility and implementation. Sensitivity analyses, discussed below, may also be an option. If affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed. The license amendment request should describe the licensee's re-analysis effort and provide statements regarding the acceptability of the proposed implementation, including modifications, against each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 of this guide.

1.3.2 The NRC staff has performed an evaluation of the Conforms- There are impact of the AST on three representative operating no plant modifications reactors (Ref. 14). This evaluation determined that that are planned to radiological analysis results based on the TID-14844 implement the AST source term assumptions (Ref. 1) and the whole body analyses. The and thyroid methodology generally bound the results radiological and from analyses based on the AST and TEDE nonradiological impacts methodology. Licensees may use the applicable of full scope conclusions of this evaluation in addressing the impact implementation of the of the AST on design basis radiological analyses. AST have been However, this does not exempt the licensee from considered and evaluating the remaining radiological and discussed in the nonradiological impacts of the AST implementation and License Amendment the impacts of the associated plant modifications. For Request, as applicable.

example, a selective implementation based on the timing insights of the AST may change the required isolation time for the containment purge dampers from 2.5 seconds to 5.0 seconds. This application might be acceptable without dose calculations. However, evaluations may need to be performed regarding the ability of the damper to close against increased containment pressure or the ability of ductwork downstream of the dampers to withstand increased stresses.

1.3.2 For full implementation, a complete DBA LOCA analysis Conforms - The DBA as described in Appendix A of this guide should be LOCA analysis is performed, as a minimum. Other design basis analyses provided in this License are updated in accordance with the guidance in this Amendment Request section. which is consistent with Appendix A.

A4-7 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 1.3.2 A selective implementation of an AST and any Not Applicable - This associated facility modification based on the AST should License Amendment evaluate all the radiological and nonradiological impacts Request is a full scope of the proposed actions as they apply to the particular AST implementation implementation. Design basis analyses are updated in that evaluates the dose accordance with the guidance in this section. There is consequences of the no minimum requirement that a DBA LOCA analysis be most severe VEGP performed. The analyses performed need to address all DBAs.

impacts of the proposed modification, the selected characteristics of the AST, and if dose calculations are performed, the TEDE criteria. For selective implementations based on the timing characteristic of the AST, e.g., change in the closure timing of a containment isolation valve, re-analysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident initiation and the onset of the gap release phase. Longer time delays may be considered on an individual basis. For longer time delays, evaluation of the radiological consequences and other impacts of the delay, such as blockage by debris in sump water, may be necessary. If affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed.

1.3.3 It may be possible to demonstrate by sensitivity or Not Applicable- The scoping evaluations that existing analyses have VEGP AST analysis sufficient margin and need not be recalculated. As used does not rely on in this guide, a sensitivity analysis is an evaluation that sensitivity or scoping considers how the overall results vary as an input analyses.

parameter (in this case, AST characteristics) is varied. A scoping analysis is a brief evaluation that uses conservative, simple methods to show that the results of the analysis bound those obtainable from a more complete treatment. Sensitivity analyses are particularly applicable to suites of calculations that address diverse components or plant areas but are otherwise largely based on generic assumptions and inputs. Such cases might include post accident vital area access dose calculations, shielding calculations, and equipment environmental qualification (integrated dose). It may be possible to identify a bounding case, re-analyze that case, and use the results to draw conclusions regarding the remainder of the analyses. It may also be possible to show that for some analyses the whole body and thyroid doses determined with the previous source term A4-8 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis would bound the TEDE obtained using the AST. Where present, arbitrary "designer margins" may be adequate to bound any impact of the AST and TEDE criteria. If sensitivity or scoping analyses are used, the license amendment request should include a discussion of the analyses performed and the conclusions drawn.

Scoping or sensitivity analyses should not constitute a significant part of the evaluations for the design basis exclusion area boundary (EAB), low population zone (LPZ), or control room dose.

1.3.4 Full implementation of the AST replaces the previous Not Applicable- The accident source term with the approved AST and the VEGP AST design TEDE criteria for all design basis radiological analyses. basis radiological The implementation may have been supported in part analyses do not rely on by sensitivity or scoping analyses that concluded many sensitivity or scoping of the design basis radiological analyses would remain analyses.

bounding for the AST and the TEDE criteria and would not require updating. After the implementation is complete, there may be a subsequent need (e.g., a planned facility modification) to revise these analyses or to perform new analyses. For these recalculations, the NRC staff expects that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as-needed basis. Re-evaluation using the previously approved source term may not be appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the facility, use of the AST and TEDE criteria in new applications at the facility do not constitute a change in analysis methodology that would require NRC approval.7 1.3.4 This guidance is also applicable to selective Not Applicable - This is implementations to the extent that the affected analyses a full scope License are within the scope of the approved implementation as Amendment Request described in the facility design basis. In these cases, the that evaluates the dose characteristics of the AST and TEDE criteria identified in consequences of the the facility design basis need to be considered in most severe VEGP updating the analyses. Use of other characteristics of DBAs.

the AST or TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, requires prior NRC staff approval under 10 CFR 50.67.

1.3.5 Current environmental qualification (EQ) analyses may Conforms - The VEGP be impacted by a proposed plant modification AST License associated with the AST implementation. The EQ Amendment Request is A4-9 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis analyses that have assumptions or inputs affected by not proposing to modify the plant modification should be updated to address the equipment these impacts. The NRC staff is assessing the effect of qualification design increased cesium releases on EQ doses to determine basis to adopt AST.

whether licensee action is warranted. Until such time as The VEGP EQ analysis this generic issue is resolved, licensees may use either will continue to be the AST or the TID14844 assumptions for performing based on TID-14844 the required EQ analyses. However, no plant assumptions.

modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue. The EQ dose estimates should be calculated using the design basis survivability period.

1.4 The use of an AST changes only the regulatory Not Applicable - No assumptions regarding the analytical treatment of the facility modifications are design basis accidents. The AST has no direct effect on proposed or planned as the probability of the accident. Use of an AST alone implementation actions cannot increase the core damage frequency (CDF) or of the FHA AST the large early release frequency (LERF). However, analysis.

facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facility's PRA, the impact on the existing PRAs should be evaluated.

1.4 Consideration should be given to the risk impact of Not Applicable- The proposed implementations that seek to remove or VEGP AST License downgrade the performance of previously required Amendment Request is engineered safeguards equipment on the basis of the not seeking to remove reduced postulated doses. The NRC staff may request or downgrade the risk information if there is a reason to question adequate performance of protection of public health and safety. previously required engineered safeguards equipment on the basis of the reduced postulated doses.

1.4 The licensee may elect to use risk insights in support of Not Applicable- The proposed changes to the design basis that are not VEGP AST License addressed in currently approved NRC staff positions. Amendment Request is For guidance, refer to Regulatory Guide 1.174, "An not utilizing risk insights Approach for Using Probabilistic Risk Assessment in as a basis for any Risk-Informed Decisions on Plant-Specific Changes to proposed changes.

the Licensing Basis" (Ref. 15).

1.5 According to 10 CFR 50.90, an application for an Conforms- The License amendment must fully describe the changes desired Amendment Request is and should follow, as far as applicable, the form formatted in A4-10 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis prescribed for original applications. Regulatory Guide accordance with 1.70, "Standard Format and Content of Safety Analysis accepted NRC/industry Reports for Nuclear Power Plants (LWR Edition)" (Ref guidance. The request 16), provides additional guidance. The NRC staff's describes the finding that the amendment may be approved must be radiological and based on the licensee's analyses, since it is these nonradiological impacts analyses that will become part of the design basis of the of the VEGP AST facility. The amendment request should describe the analysis. Consistent licensee's analyses of the radiological and with previous nonradiological impacts of the proposed modification in precedent, affected sufficient detail to support review by the NRC staff. The FSAR pages are not staff recommends that licensees submit affected FSAR included in the pages annotated with changes that reflect the revised analyses. However, a analyses or submit the actual calculation detailed summary of documentation. the AST dose calculations are included. Approval of this License Amendment Request will result in the necessary revisions to the FSAR, with revised FSAR pages submitted pursuant to 10 CFR 50.71(e).

1.5 If the licensee has used a current approved version of DBA dose analyses an NRC-sponsored computer code, the NRC staff contained in this LAR review can be made more efficient if the licensee submittal utilize Sercos identifies the code used and submits the inputs that the version of RADTRAD licensee used in the calculations made with that code. In Version 3.10.

many cases, this will reduce the need for NRC staff confirmatory analyses. This recommendation does not constitute a requirement that the licensee use NRC-sponsored computer codes.

A4-11 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 1.6 Requirements for updating the facility's final safety Conforms- Approval of analysis report (FSAR) are in 10 CFR 50.71, this License "Maintenance of Records, Making of Reports." The Amendment Request regulations in 10 CFR 50.71(e) require that the FSAR will result in the be updated to include all changes made in the facility or necessary revisions to procedures described in the FSAR and all safety the FSAR, with revised evaluations performed by the licensee in support of FSAR pages submitted requests for license amendments or in support of pursuant to 10 CFR conclusions that changes did not involve unreviewed 50.71(e).

safety questions. The analyses required by 10 CFR 50.67 are subject to this requirement. The affected radiological analysis descriptions in the FSAR should be updated to reflect the replacement of the design basis source term by the AST. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric results. Regulatory Guide 1.70 (Ref.

16) provides additional guidance. The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis.

2.1 The AST must be based on major accidents, Conforms- This License hypothesized for the purposes of design analyses or Amendment Request consideration of possible accidental events, that could applies the AST result in hazards not exceeded by those from other methods when accidents considered credible. The AST must address evaluating the dose events that involve a substantial meltdown of the core consequences of the with the subsequent release of appreciable quantities of most severe DBAs fission products. applicable to VEGP.

2.2 The AST must be expressed in terms of times and rates Conforms - For the of appearance of radioactive fission products released DBAs that release to into containment, the types and quantities of the Containment (LOCA radioactive species released, and the chemical forms of and Control Rod iodine released. Ejection), the AST is expressed in terms of times and rates of release of radioactive fission products, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

2.3 The AST must not be based upon a single accident Conforms- This License scenario but instead must represent a spectrum of Amendment Request credible severe accident events. Risk insights may be considers a number of used, not to select a single risk-significant accident, but release scenarios, as rather to establish the range of events to be considered. applicable, for the A4-12 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis Relevant insights from applicable severe accident DBAs being revised for research on the phenomenology of fission product use of AST. The most release and transport behavior may be considered. limiting of these releases are analyzed for radiological consequences.

2.4 The AST must have a defensible technical basis Conforms- The DBA supported by sufficient experimental and empirical data, AST dose calculations be verified and validated, and be documented in a have been developed scrutable form that facilitates public review and based on NUREG-1465 discourse. and this Regulatory Guide. The calculations, which utilize RADTRAD were developed in accordance with 10 CFR 50 Appendix B, Criterion III.

2.5 The AST must be peer-reviewed by appropriately Conforms- The VEGP qualified subject matter experts. The peer-review AST dose calculations comments and their resolution should be part of the have been developed documentation supporting the AST. and independently reviewed by internal experts at SNC. SNC internal experts are qualified via SNC internal processes and procedures in the performance of dose analyses. The calculations were developed in accordance with 10 CFR 50 Appendix B program, Criterion III.

A4-13 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 3.1 The inventory of fission products in the reactor core and Conforms -The VEGP available for release to the containment should be DBAs that release to based on the maximum full power operation of the core the Containment are with, as a minimum, current licensed values for fuel the LOCA, FHA, and enrichment, fuel burnup, and an assumed core power Control Rod Ejection.

equal to the current licensed rated thermal power times Core Inventory has the ECCS evaluation uncertainty. The period of been determined using irradiation should be of sufficient duration to allow the an appropriate isotope activity of dose-significant radionuclides to reach generation and equilibrium or to reach maximum values. The core depletion code, such as inventory should be determined using an appropriate ORIGEN2 or ORIGEN-isotope generation and depletion computer code such ARP.

as ORIGEN 2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TID14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

3.1 For the DBA LOCA, all fuel assemblies in the core are With the exception of assumed to be affected and the core average inventory DBAs where cladding should be used. For DBA events that do not involve the damage is postulated entire core, the fission product inventory of each of the with a gap release, the damaged fuel rods is determined by dividing the total analyses of events core inventory by the number of fuel rods in the core. To which involve fuel account for differences in power level across the core, damage assume that radial peaking factors from the facility's core operating the entire core is limits report (COLR) or technical specifications should affected with a source be applied in determining the inventory of the damaged term based upon full rods. power, core average conditions. The source term for DBAs where cladding damage is postulated with a gap release is derived from the core source term, the number of damaged fuel rods, and a conservative assembly peaking factor, which exceeds the maximum fuel rod peaking factor specified in the COLR.

3.1 No adjustment to the fission product inventory should be The analysis of the made for events postulated to occur during power FHA considers operations at less than full rated power or those radioactive decay postulated to occur at the beginning of core life. For between the time of events postulated to occur while the facility is shutdown, core shutdown and the A4-14 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis e.g., a fuel handling accident, radioactive decay from beginning of fuel the time of shutdown may be modeled. movement.

3.2 The core inventory release fractions, by radionuclide Conforms - The LOCA groups, for the gap release and early in-vessel damage AST calculation models phases for DBA LOCAs are listed in Table 1 for BWRs Table 2 in the release and Table 2 for PWRs. These fractions are applied to fraction and timing file.

the equilibrium core inventory described in Regulatory Position 3.1.

Table 2 PWR Core Inventory Fraction Released Into Containment Gap Early Release In-vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.35 0.4 Alkali Metals 0.05 0.25 0.3 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 3.2 For non-LOCA events, the fractions of the core Conforms - The FHA, inventory assumed to be in the gap for the various Control Rod Ejection, radionuclides are given in Table 3. The release fractions and Locked Rotor from Table 3 are used in conjunction with the fission accidents result in fuel product inventory calculated with the maximum core damage, so the non-radial peaking factor. LOCA gap fractions of Table 3 are used. While Table 3.11 Non-LOCA Fraction of Fission Product the SGTR and MSLB Inventory in Gap accidents conservatively assume Table 3 a pre-existing 1%

leaking fuel source term Group Fraction for everything except I-131 0.08 noble gases and Kr-85 0.10 Iodines, this is not the Other Noble Gases 0.05 result of damage Other Halogens 0.05 caused by the accident, Alkali Metals 0.12 and so the non-LOCA gap fractions of Table 3 are not included for these events.

A4-15 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 3.3 Table 4 tabulates the onset and duration of each Conforms - The LOCA sequential release phase for DBA LOCAs at PWRs and AST calculation models BWRs. The specified onset is the time following the Table 4 in the release initiation of the accident (i.e., time = 0). The early in- fraction and timing file.

vessel phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase.12 For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 4 LOCA Release Phases (PWR)

Phase Onset Duration Gap Release 30 sec 0.5 hr Early In-vessel 0.5 hr 1.3 hr 3.3 For facilities licensed with leak-before-break Conforms - The LOCA methodology, the onset of the gap release phase may AST calculation models be assumed to be 10 minutes. A licensee may propose Table 4 in the release an alternative time for the onset of the gap release fraction and timing file.

phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

3.4 Elements listed in Table 5 in each radionuclide group Conforms The source that should be considered in design basis analyses. term in the design basis analysis represents the Table 5 most dose significant Radionuclide Groups isotopes from the Group Elements elements listed in Table Noble Gases Xe, Kr 5 of Regulatory Guide Halogens I, Br 1.183.

Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthenides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np A4-16 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 3.5 Of the radioiodine released from the reactor coolant Conforms - The system (RCS) to the containment in a postulated chemical composition accident, 95 percent of the iodine released should be of the iodine released assumed to be cesium iodide (CsI), 4.85 percent from the RCS to elemental iodine, and 0.15 percent organic iodide. This containment in the includes releases from the gap and the fuel pellets. With LOCA event is 95%

the exception of elemental and organic iodine and noble aerosol, 4.85%

gases, fission products should be assumed to be in elemental, and 0.15%

particulate form. The same chemical form is assumed in organic. All non iodine releases from fuel pins in FHAs and from releases from and non-noble gas the fuel pins through the RCS in DBAs other than FHAs fission products are or LOCAs. However, the transport of these iodine assumed to be in species following release from the fuel may affect these particulate form. The assumed fractions. The accident-specific appendices to chemical composition this regulatory guide provide additional details. of iodine species in the non-LOCA events are based upon the guidance in the respective appendices of Regulatory Guide 1.183.

3.6 The amount of fuel damage caused by non-LOCA Conforms - The amount design basis events should be analyzed to determine, of fuel damage in the for the case resulting in the highest radioactivity release, Locked Rotor event is the fraction of the fuel that reaches or exceeds the based upon the fraction initiation temperature of fuel melt and the fraction of fuel of the core which elements for which the fuel clad is breached. Although experiences DNB as the NRC staff has traditionally relied upon the departure reported in the Updated from nucleate boiling ratio (DNBR) as a fuel damage Final Safety Analysis criterion, licensees may propose other methods to the Report (FSAR). The NRC staff, such as those based upon enthalpy fraction of the fuel rods deposition, for estimating fuel damage for the purpose assumed to melt in the of establishing radioactivity releases. Control Rod Ejection event is conservatively based upon the portion of the fuel centerline that is calculated to exceed the melting temperature as documented in the FSAR.

4.1.1 The dose calculations should determine the TEDE. Conforms - The AST TEDE is the sum of the committed effective dose dose consequences are equivalent (CEDE) from inhalation and the deep dose calculated in TEDE.

equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny A4-17 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.13 4.1.2 The exposure-to-CEDE factors for inhalation of Conforms - Dose radioactive material should be derived from the data Conversion Factors for provided in ICRP Publication 30, "Limits for Intakes of inhalation in this Radionuclides by Workers" (Ref. 19). Table 2.1 of analysis are taken from Federal Guidance Report 11, "Limiting Values of Table 2.1 of Federal Radionuclide Intake and Air Concentration and Dose Guidance Report 11.

Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of persons offsite Conforms - Offsite should be assumed to be 3.5 x 10-4 cubic meters per breathing rates used in second. From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident, the the analysis are breathing rate should be assumed to be 1.8 x 10-4 cubic consistent with the meters per second. After that and until the end of the values specified in accident, the rate should be assumed to be 2.3 x 10-4 Section 4.1.3 of cubic meters per second. Regulatory Guide 1.183.

4.1.4 The DDE should be calculated assuming submergence Conforms - Dose in semi-infinite cloud assumptions with appropriate Conversion Factors for credit for attenuation by body tissue. The DDE is air submergence are nominally equivalent to the effective dose equivalent taken from the Table (EDE) from external exposure if the whole body is III.1 of Federal irradiated uniformly. Since this is a reasonable Guidance Report 12.

assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table III.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21),

provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

4.1.5 The TEDE should be determined for the most limiting Conforms - The TEDE person at the EAB. The maximum EAB TEDE for any was determined for the two-hour period following the start of the radioactivity most limiting person at release should be determined and used in determining the EAB. The maximum compliance with the dose criteria in 10 CFR 50.67. The two-hour TEDE was maximum two-hour TEDE should be determined by determined by calculating the postulated dose for a series of small time calculating the increments and performing a "sliding" sum over the postulated dose for a increments for successive two-hour periods. The series of small time maximum TEDE obtained is submitted. The time increments and A4-18 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis increments should appropriately reflect the progression performing a 'sliding' of the accident to capture the peak dose interval sum over increments between the start of the event and the end of for successive two-hour radioactivity release (see also Table 6). periods.

4.1.6 TEDE should be determined for the most limiting Conforms - The TEDE receptor at the outer boundary of the low population is determined for the zone (LPZ) and should be used in determining most limiting person at compliance with the dose criteria in 10 CFR 50.67. the LPZ.

4.1.7 No correction should be made for depletion of the Conforms - No effluent plume by deposition on the ground. correction is made for deposition of the effluent plume by deposition on the ground.

4.2.1 The TEDE analysis should consider all sources of Conforms - The radiation that will cause exposure to control room analyses consider the personnel. The applicable sources will vary from facility applicable sources of to facility, but typically will include: contamination to the Contamination of the control room atmosphere control room by the intake or infiltration of the radioactive atmosphere for each material contained in the radioactive plume event.

released from the facility, Contamination of the control room atmosphere With respect to external by the intake or infiltration of airborne radioactive and containment shine material from areas and structures adjacent to sources and their the control room envelope, impact on control room Radiation shine from the external radioactive doses, the physical plume released from the facility, design of the control Radiation shine from radioactive material in the room envelope and the reactor containment, surrounding auxiliary Radiation shine from radioactive material in building provide more systems and components inside or external to than 18" of concrete the control room envelope, e.g., radioactive shielding between the material buildup in recirculation filters. operators and shine sources in all directions around the control room.

The Control Room Emergency Filtration System filters are located outside of and above the control room envelope. The control room ceiling is approximately 18" thick.

Accordingly, shielding A4-19 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis from the walls and the filter unit casings prevents an appreciable dose to the operators during the accident.

4.2.2 The radioactive material releases and radiation levels Conforms - The SNC used in the control room dose analysis should be AST dose calculations determined using the same source term, transport, and use the same source release assumptions used for determining the EAB and term, transport, and the LPZ TEDE values, unless these assumptions would release assumptions for result in non-conservative results for the control room. Control Room, EAB, and LPZ dose values.

4.2.3 The models used to transport radioactive material into Conforms - The models and through the control room,15 and the shielding used to transport models used to determine radiation dose rates from radioactive material into external sources, should be structured to provide and through the control suitably conservative estimates of the exposure to room have been control room personnel. structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate Conforms - For the airborne radioactive material within the control room AST DBAs covered may be assumed. Such features may include control under this License room isolation or pressurization, or intake or Amendment Request, recirculation filtration. Refer to Section 6.5.1, "ESF credit is taken for Atmospheric Cleanup System," of the SRP (Ref. 3) and control room isolation Regulatory Guide 1.52, "Design, Testing, and and reconfiguring into Maintenance Criteria for Post accident Engineered- the emergency Safety-Feature Atmosphere Cleanup System Air ventilation mode upon Filtration and Adsorption Units of Light-Water-Cooled accident initiation by a Nuclear Power Plants" (Ref. 25), for guidance. The high radiation or Safety control room design is often optimized for the DBA Injection signal, where LOCA and the protection afforded for other accident appropriate.

sequences may not be as advantageous. In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents.

Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the A4-20 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis effects of different radionuclide accident isotopic mixes on monitor response.

4.2.5 Credit should generally not be taken for the use of Conforms- No credit is personal protective equipment or prophylactic drugs. taken for the use of Deviations may be considered on a case-by-case basis. personal protective equipment or prophylactic drugs.

4.2.6 The dose receptor for these analyses is the hypothetical Conforms - Control maximum exposed individual who is present in the room occupancy and control room for 100% of the time during the first 24 breathing rates are hours after the event, 60% of the time between 1 and 4 consistent with this days, and 40% of the time from 4 days to 30 days.16 regulatory position.

For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10-4 cubic meters per second.

4.2.7 Control room doses should be calculated using dose Conforms - Control conversion factors identified in Regulatory Position 4.1 room doses are above for use in offsite dose analyses. The DDE from calculated using dose photons may be corrected for the difference between conversion factors finite cloud geometry in the control room and the semi- identified in Position 4.1 infinite cloud assumption used in calculating the dose above.

conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDE, to a finite cloud dose, DDEfinite , where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

DDE V 0.338 DDEfinite=

1173 4.3 The guidance provided in Regulatory Positions 4.1 and Not Applicable - This 4.2 should be used, as applicable, in re-assessing the full scope AST radiological analyses identified in Regulatory Position implementation LAR is 1.3.1, such as those in NUREG-0737 (Ref. 2). Design for the radiological envelope source terms provided in NUREG-0737 should consequences of major be updated for consistency with the AST. In general, VEGP DBAs.

radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.

4.4 The radiological criteria for the EAB, the outer boundary The EAB and LPZ of the LPZ, and for the control room are in 10 CFR acceptance criteria 50.67. These criteria are stated for evaluating reactor from Table 6 of RG accidents of exceedingly low probability of occurrence 1.183 are applied. The and low risk of public exposure to radiation, e.g., a control room acceptance of 5 rem A4-21 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis large-break LOCA. The control room criterion applies to TEDE is taken from 10 all accidents. For events with a higher probability of CFR 50.67(b)(2)(iii).

occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.

Table 6 Accident Dose Criteria Accident or EAB and Analysis Release Case LPZ Dose Duration Criteria LOCA 25 rem 30 days for TEDE containment and ECCS leakage PWR Steam Affected SG: time to Generator Tube isolate; Unaffected Rupture SG(s): until cold shutdown is established Fuel Damage 25 rem or Pre-incident TEDE Spike Coincident 2.5 rem Iodine Spike TEDE PWR Main Until cold shutdown Steam Line is established Break Fuel Damage 25 rem or Pre-incident TEDE Spike Coincident 2.5 rem Iodine Spike TEDE PWR Locked 2.5 rem Until cold shutdown Rotor Accident TEDE is established PWR Rod 6.3 rem 30 days for Ejection TEDE containment Accident pathway; until cold shutdown is established for secondary pathway Fuel Handling 6.3 rem 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Accident TEDE A4-22 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis The column labeled Analysis Release Duration is a summary of the assumed radioactivity release durations identified in the individual appendices to this guide.

Refer to these appendices for complete descriptions of the release pathways and durations.

4.4 The acceptance criteria for the various NUREG-0737 Conforms - The EAB (Ref. 2) items generally reference General Design and LPZ acceptance Criteria 19 (GDC 19) from Appendix A to 10 CFR Part criteria from Table 6 of 50 or specify criteria derived from GDC 19. These RG 1.183 are applied.

criteria are generally specified in terms of whole body The control room dose, or its equivalent to any body organ. For facilities occupant acceptance applying for, or having received, approval for the use of criteria of 5 rem TEDE an AST, the applicable criteria should be updated for is taken from 10 CFR consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

50.67(b)(2)(iii).

5.1.1 The evaluations required by 10 CFR 50.67 are re- Conforms- The VEGP analyses of the design basis safety analyses and AST dose calculations evaluations required by 10 CFR 50.34; they are were prepared, considered to be a significant input to the evaluations reviewed, and required by 10 CFR 50.92 or 10 CFR 50.59. These maintained, by SNC analyses should be prepared, reviewed, and maintained under a 10 CFR 50 in accordance with quality assurance programs that Appendix B Quality comply with Appendix B, "Quality Assurance Criteria for Assurance program.

Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

5.1.1 These design basis analyses were structured to provide Not Applicable- This a conservative set of assumptions to test the License Amendment performance of one or more aspects of the facility Request is not design. Many physical processes and phenomena are proposing deviations to represented by conservative, bounding assumptions conformance with this rather than being modeled directly. The staff has Regulatory Guide.

selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion. Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence -- the proposed deviation may not be conservative for other accident sequences.

5.1.2 Credit may be taken for accident mitigation features that Conforms - Only safety-are classified as safety-related, are required to be related Engineered operable by technical specifications, are powered by Safety Features are emergency power sources, and are either automatically credited in the analysis A4-23 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis actuated or, in limited cases, have actuation with an assumed single requirements explicitly addressed in emergency active failure that operating procedures. The single active component results in the greatest failure that results in the most limiting radiological impact on the consequences should be assumed. Assumptions radiological regarding the occurrence and timing of a loss of offsite consequences. A loss power should be selected with the objective of of offsite power is maximizing the postulated radiological consequences. assumed concurrent with the start of each event as that maximizes the dose impact.

5.1.3 The numeric values that are chosen as inputs to the Conforms - Numerical analyses required by 10 CFR 50.67 should be selected values are selected and with the objective of determining a conservative biased for each postulated dose. In some instances, a particular application in a parameter may be conservative in one portion of an conservative direction analysis but be nonconservative in another portion of with the objective of the same analysis. For example, assuming minimum maximizing the dose containment system spray flow is usually conservative consequences.

for estimating iodine scrubbing, but in many cases may Numerical values for be nonconservative when determining sump pH. parameters which are Sensitivity analyses may be needed to determine the controlled by Technical appropriate value to use. As a conservative alternative, Specifications are the limiting value applicable to each portion of the either used as direct analysis may be used in the evaluation of that portion. A inputs in the analysis, single value may not be applicable for a parameter for or more conservative the duration of the event, particularly for parameters values may be used to affected by changes in density. For parameters enhance safety margin.

addressed by technical specifications, the value used in the analysis should be that specified in the technical specifications.18 If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing, e.g., steam generator nondestructive testing (NDT), consideration should be given to the degradation that may occur between periodic tests in establishing the analysis value.

5.1.4 The NRC staff considers the implementation of an AST Conforms- The VEGP to be a significant change to the design basis of the DBA analysis facility that is voluntarily initiated by the licensee. In assumptions and order to issue a license amendment authorizing the use methods are of an AST and the TEDE dose criteria, the NRC staff compatible with the must make a current finding of compliance with AST and the TEDE regulations applicable to the amendment. The criteria.

characteristics of the ASTs and the revised dose A4-24 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses.

The NRC staff may find that new or unreviewed issues are created by a particular site-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109, "Backfitting." However, prior design bases that are unrelated to the use of the AST, or are unaffected by the AST, may continue as the facility's design basis. Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.

5.2 The appendices to this regulatory guide provide Conforms - See Tables accident-specific assumptions that are acceptable to the B, C, D, E, F, and G of staff for performing analyses that are required by 10 this Enclosure.

CFR 50.67. The DBAs addressed in these attachments were selected from accidents that may involve damage to irradiated fuel. This guide does not address DBAs with radiological consequences based on technical specification reactor or secondary coolant-specific activities only. The inclusion or exclusion of a particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is required or not required.

Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

5.2 The NRC staff has determined that the analysis Conforms - See Tables assumptions in the appendices to this guide provide an B, C, D, E, F, and G of integrated approach to performing the individual this Enclosure.

analyses and generally expects licensees to address each assumption or propose acceptable alternatives.

Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration. The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency.

A4-25 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis 5.2 The NRC is committed to using probabilistic risk Conforms- PRA was analysis (PRA) insights in its regulatory activities and not used as a basis for will consider licensee proposals for changes in analysis acceptability of this assumptions based upon risk insights. The staff will not AST License approve proposals that would reduce the defense in Amendment Request.

depth deemed necessary to provide adequate protection for public health and safety. In some cases, this defense in depth compensates for uncertainties in the PRA analyses and addresses accident considerations not adequately addressed by the core damage frequency (CDF) and large early release frequency (LERF) surrogate indicators of overall risk.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the Conforms - The X/Q LPZ, and the control room that were approved by the used for the EAB and staff during initial facility licensing or in subsequent the LPZ were licensing proceedings may be used in performing the previously approved by radiological analyses identified by this guide. the NRC.

Methodologies that have been used for determining X/Q values are documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28).

5.3 References 22 and 28 should be used if the FSAR X/Q Conforms - The onsite values are to be revised or if values are to be X/Q values are either determined for new release points or receptor distances. those already used in Fumigation should be considered where applicable for the FSAR or were the EAB and LPZ. For the EAB, the assumed fumigation developed based on period should be timed to be included in the worst 2- the approved hour exposure period. The NRC computer code PAVAN meteorological data.

(Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) ARCON96 (Reference and its use is acceptable to the NRC staff. The 26) was used to methodology of the NRC computer code ARCON9619 generate revised onsite (Ref. 26) is generally acceptable to the NRC staff for dose-receptor pairs.

use in determining control room X/Q values.

Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

6.0 The assumptions in Appendix I to this guide are Conforms - VEGP is acceptable to the NRC staff for performing radiological retaining the use of the A4-26 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis assessments associated with equipment qualification. TID 14844 source term The assumptions in Appendix I will supersede as the basis for Regulatory Positions 2.c(1) and 2.c(2) and Appendix D Environmental of Revision 1 of Regulatory Guide 1.89, "Environmental Qualification.

Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (Ref. 11), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.

The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue.

Footnote For example, a proposed modification to change the Conforms - No 6 timing of a containment isolation valve from 2.5 seconds modifications are being to 5.0 seconds might be acceptable without any dose proposed as part of this calculations. However, a proposed modification that AST License would delay containment spray actuation could involve Amendment Request.

recalculation of DBA LOCA doses, re-assessment of the containment pressure and temperature transient, recalculation of sump pH, re-assessment of the emergency diesel generator loading sequence, integrated doses to equipment in the containment, and more.

Footnote In performing screenings and evaluations pursuant to 10 Not Applicable - This 7 CFR 50.59, it may be necessary to compare dose activity is a License results expressed in terms of whole body and thyroid Amendment Request with new results expressed in terms of TEDE. In these made pursuant to 10 cases, the previous thyroid dose should be multiplied by CFR Part 90.

0.03 and the product added to the whole body dose.

The result is then compared to the TEDE result in the screenings and evaluations. This change in dose methodology is not considered a change in the method of evaluation if the licensee was previously authorized to use an AST and the TEDE criteria under 10 CFR 50.67.

Footnote The uncertainty factor used in determining the core Conforms - A 1.02 8 inventory should be that value provided in Appendix K to uncertainty factor is 10 CFR Part 50, typically 1.02. used for those events A4-27 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis resulting in fuel damage.

Footnote Note that for some radionuclides, such as Cs-137, Conforms - A 9 equilibrium will not be reached prior to fuel offload. conservative core factor Thus, the maximum inventory at the end of life should is applied to the be used. principal radionuclides to account for cycle-to-cycle variations.

Footnote The release fractions listed here have been determined Conforms - Burnup 10 to be acceptable for use with currently approved LWR does not exceed fuel with a peak burnup up to 62,000 MWD/MTU. The 62,000 MWD/MTU at data in this section may not be applicable to cores VEGP.

containing mixed oxide (MOX) fuel.

Footnote The release fractions listed here have been determined Exception - SNC has 11 to be acceptable for use with currently approved LWR requested an exception fuel with a peak burnup up to 62,000 MWD/MTU for the LHGR limit to be provided that the maximum linear heat generation rate 7.5 kW/ft for 40% of the does not exceed 6.3 kw/ft peak rod average power for rods in an assembly.

burnups exceeding 54 GWD/MTU. As an alternative, Only the FHA is fission gas release calculations performed using NRC- impacted by this approved methodologies may be considered on a case- request. The gap by-case basis. To be acceptable, these calculations fractions used for the must use a projected power history that will bound the FHA are consistent with limiting projected plant-specific power history for the those contained in specific fuel load. For the BWR rod drop accident the PNNL-18212 Rev. 1 PWR rod ejection accident, the gap fractions are Table 2.9.

assumed to be 10% for iodines and noble gases.

Footnote In lieu of treating the release in a linear ramp manner, Conforms -RADTRAD 12 the activity for each phase can be modeled as being can model the release released instantaneously at the start of that release either in a linear ramp phase, i.e., in step increases. manner, or instantaneous release, as required.

Footnote The prior practice of basing inhalation exposure on only Conforms - Offsite 13 radioiodine and not including radioiodine in external inhalation doses are exposure calculations is not consistent with the calculated consistent definition of TEDE and the characteristics of the revised with the definition of source term. TEDE.

Footnote With regard to the EAB TEDE, the maximum two-hour Not Applicable - This 14 value is the basis for screening and evaluation under 10 activity is a License CFR 50.59. Changes to doses outside of the two-hour Amendment Request window are only considered in the context of their made pursuant to 10 impact on the maximum two-hour EAB TEDE. CFR Part 90.

Footnote The iodine protection factor (IPF) methodology of Conforms - The iodine 15 Reference 22 may not be adequately conservative for protection factor all DBAs and control room arrangements since it models methodology of a steady-state control room condition. Since many A4-28 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis analysis parameters change over the duration of the Reference 22 is not event, the IPF methodology should only be used with used in this application.

caution. The NRC computer codes HABIT (Ref. 23) and RADTRAD (Ref. 24) incorporate suitable methodologies.

Footnote This occupancy is modeled in the X/Q values Conforms - The control 16 determined in Reference 22 and should not be credited room occupancy twice. The ARCON96 Code (Ref. 26) does not assumptions are incorporate these occupancy assumptions, making it incorporated in the necessary to apply this correction in the dose dose calculations calculations.

Footnote For PWRs with steam generator alternative repair Conforms - Refer to 17 criteria, different dose criteria may apply to steam ARC line items in generator tube rupture and main steam line break Tables D and E.

analyses.

Footnote Note that for some parameters, the technical Conforms - Filter 18 specification value may be adjusted for analysis efficiencies for the purposes by factors provided in other regulatory PPAFES, the Control guidance. For example, ESF filter efficiencies are based Room (CR) on the guidance in Regulatory Guide 1.52 (Ref. 25) and Pressurization Intake in Generic Letter 99-02 (Ref. 27) rather than the Filters, and the Control surveillance test criteria in the technical specifications. Room Recirculation Generally, these adjustments address potential changes Filters are developed in the parameter between scheduled surveillance tests. from Technical Specification Surveillance requirements, with margin added for filter inefficiency and bypass leakage around the filter (in accordance with the prior CLB analyses of this type:

double the penetration allowed by TS 5.5.11 and further reduce efficiency by a 0.5%

bypass amount).

This methodology assures compliance with Technical Specification 5.5.11 requirements and US NRC RG-1.52.

A4-29 to Enclosure Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position VEGP Analysis Footnote The ARCON96 computer code contains processing Conforms - The 19 options that may yield X/Q values that are not ARCON96 processing sufficiently conservative for use in accident options and input consequence assessments or may be incompatible with parameters were based release point and ventilation intake configurations at on the release point particular sites. The applicability of these options and and ventilation intake associated input parameters should be evaluated on a configurations at case-by-case basis. The assumptions made in the VEGP.

examples in the ARCON96 documentation are illustrative only and do not imply NRC staff acceptance of the methods or data used in the example.

A4-30 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis A-1 Acceptable assumptions regarding core Conforms - See inventory and the release of radionuclides discussions in Table A.

from the fuel are provided in Regulatory Position 3 of this guide.

A-2 If the sump or suppression pool pH is Conforms - The pH of the controlled at values of 7 or greater, the containment sump is chemical form of radioiodine released to the maintained equal to or containment should be assumed to be 95% greater than 7.0 after the cesium iodide (CsI), 4.85 percent elemental onset of the spray iodine, and 0.15 percent organic iodide. recirculation mode.

Iodine species, including those from iodine re- Therefore, the radioiodine evolution, for sump or suppression pool pH composition of 95 percent values less than 7 will be evaluated on a cesium iodide, 4.85 case-by-case basis. Evaluations of pH should percent elemental iodine, consider the effect of acids and bases created and 0.15 percent organic during the LOCA event, e.g., radiolysis iodide is used. The products. With the exception of elemental and containment sump pH has organic iodine and noble gases, fission been evaluated for the products should be assumed to be in impact of the alternate particulate form. source term and confirms that the sump pH remains greater than 7.0. In addition, VEGP uses trisodium phosphate to create a buffered sump solution that is resistant to change in pH.

A-3.1 The radioactivity released from the fuel should Conforms - The be assumed to mix instantaneously and radioactivity released from homogeneously throughout the free air the fuel is modeled as volume of the primary containment in PWRs mixing instantaneously or the drywell in BWRs as it is released. This and homogeneously in the distribution should be adjusted if there are Containment.

internal compartments that have limited ventilation exchange. The suppression pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel phase.

A-3.2 Reduction in airborne radioactivity in the Conforms - An aerosol containment by natural deposition within the natural deposition rate of A4-31 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis containment may be credited. Acceptable 0.1 hr-1 is assumed based models for removal of iodine and aerosols are upon values presented described in Chapter 6.5.2, Containment Section VI of NUREG/CR-Spray as a Fission Product Cleanup System, 6189.

of the Standard Review Plan (SRP), NUREG-0800 (Ref. A-1) and in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3).

The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.

A-3.3 Reduction in airborne radioactivity in the Conforms - Containment containment by containment spray systems Spray is credited for that have been designed and are maintained elemental iodine and in accordance with Chapter 6.5.2 of the SRP aerosol removal.

(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). This simplified model is incorporated into the analysis code RADTRAD (Refs. A-1 to A-3).

A-3.3 The evaluation of the containment sprays Conforms - Containment should address areas within the primary Spray covers less than containment that are not covered by the spray 90% of the Containment drops. The mixing rate attributed to natural volume, so the modeling convection between sprayed and unsprayed includes both the sprayed regions of the containment building, provided volume and unsprayed that adequate flow exists between these volume. A flow rate of regions, is assumed to be two turnovers of the 21,000 cfm is used unsprayed regions per hour, unless other between the sprayed and rates are justified. The containment building unsprayed volume which atmosphere may be considered a single, well- correlates to two turnovers mixed volume if the spray covers at least 90% of the unsprayed region of the volume and if adequate mixing of per hour.

unsprayed compartments can be shown.

A-3.3 The SRP sets forth a maximum Conforms - Elemental and decontamination factor (DF) for elemental aerosol removal A4-32 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis iodine based on the maximum iodine activity coefficients are calculated in the primary containment atmosphere when for the sprayed regions of the sprays actuate, divided by the activity of the containment using the iodine remaining at some time after guidelines of Chapter decontamination. The SRP also states that 6.5.2 of the Standard the particulate iodine removal rate should be Review Plan. The reduced by a factor of 10 when a DF of 50 is elemental iodine removal reached. The reduction in the removal rate is coefficients are limited to not required if the removal rate is based on a maximum value of the calculated time-dependent airborne 13.7/hr, and are set to aerosol mass. There is no specified maximum zero when the elemental DF for aerosol removal by sprays. The iodine decontamination maximum activity to be used in determining factor (DF) reaches a the DF is defined as the iodine activity in the value of 200. The aerosol columns labeled Total in Tables 1 and 2 of removal coefficients are this guide multiplied by 0.05 for elemental reduced by a factor of 10 iodine and by 0.95 for particulate iodine (i.e., when the aerosol DF aerosol treated as particulate in SRP reaches 50.

methodology).

A-3.4 Reduction in airborne radioactivity in the Conforms - No credit is containment by in-containment recirculation taken for in-containment filter systems may be credited if these recirculation filter systems.

systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs.

A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

A-3.5 Reduction in airborne radioactivity in the Not Applicable - VEGP is containment by suppression pool scrubbing in a PWR.

BWRs should generally not be credited.

However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7).

Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

A4-33 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis A-3.6 Reduction in airborne radioactivity in the Conforms - No credit is containment by retention in ice condensers, or taken for ice condensers other engineering safety features not or other engineering addressed above, should be evaluated on an safety features to reduce individual case basis. See Section 6.5.4 of the airborne radioactivity in SRP (Ref. A-1). containment.

A-3.7 The primary containment (i.e., drywell for Conforms - The Mark I and II containment designs) should be containment leak rate for assumed to leak at the peak pressure the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is the technical specification leak rate for the first 24 maximum value allowed hours. For PWRs, the leak rate may be by the VEGP Technical reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the Specifications. It is technical specification leak rate. For BWRs, reduced to 50% of that leakage may be reduced after the first 24 value after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . An hours, if supported by plant configuration and additional 5% margin is analyses, to a value not less than 50% of the applied to the leak rate technical specification leak rate. Leakage values for conservatism.

from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

A-3.7 For BWRs with Mark III containments, the Not Applicable. VEGP is a leakage from the drywell into the primary PWR.

containment should be based on the steaming rate of the heated reactor core, with no credit for core debris relocation. This leakage should be assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment.

A-3.8 If the primary containment is routinely purged Conforms - Based upon during power operations, releases via the the isolation of the mini-purge system prior to containment isolation purge flow within 30 should be analyzed and the resulting doses seconds, the mini-purge summed with the postulated doses from other system will be isolated release paths. The purge release evaluation before the onset of the should assume that 100% of the radionuclide gap release as defined in inventory in the reactor coolant system liquid Table 4 of this Regulatory is released to the containment at the initiation Guide. Therefore, only A4-34 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis of the LOCA. This inventory should be based those nuclides in the RCS on the technical specification reactor coolant source term are available system equilibrium activity. Iodine spikes for release.

need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

A-4 For facilities with dual containment systems, Not Applicable. VEGP the acceptable assumptions related to the does not have a dual transport, reduction, and release of containment.

radioactive material in and from the secondary containment or enclosure buildings are as follows.

A-5.1 With the exception of noble gases, all the Conforms - With the fission products released from the fuel to the exception of noble gases, containment (as defined in Tables 1 and 2 of all the fission products this guide) should be assumed to released from the fuel to instantaneously and homogeneously mix in the containment is the primary containment sump water (in assumed to PWRs) or suppression pool (in BWRs) at the instantaneously and time of release from the core. In lieu of this homogeneously mix in the deterministic approach, suitably conservative primary sump water.

mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

A-5.2 The leakage should be taken as two times the The VEGP Technical sum of the simultaneous leakage from all Specifications do not components in the ESF recirculation systems provide a specific limit for above which the technical specifications, or operational leakage from licensee commitments to item III.D.1.1 of ECCS systems. However, NUREG-0737 (Ref. A-8), would require administrative limits declaring such systems inoperable. The ensure that the leakage should be assumed to start at the operational leakage earliest time the recirculation flow occurs in outside of containment these systems and end at the latest time the from ECCS systems is releases from these systems are terminated. limited to no more than 2 A4-35 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis Consideration should also be given to design gpm total, which is leakage through valves isolating ESF multiplied by two, recirculation systems from tanks vented to consistent with this atmosphere, e.g., emergency core cooling Regulatory Position. In system (ECCS) pump miniflow return to the addition, two times the refueling water storage tank. assumed leak rate of 7.0 gpm past valves that isolate return flow to the Refueling Water Storage Tank (RWST) is evaluated separately. The leakage is assumed to start at the earliest time that recirculation occurs in the ECCS system and continues for the 30-day duration of the event.

A-5.3 With the exception of iodine, all radioactive Conforms - With the materials in the recirculating liquid should be exception of iodine, all assumed to be retained in the liquid phase. radioactive materials in the recirculating liquid is modeled as being retained in the liquid phase.

A-5.4 If the temperature of the leakage exceeds Conforms - It is assumed 212°F, the fraction of total iodine in the liquid for the case when the that becomes airborne should be assumed temperature of the ECCS equal to the fraction of the leakage that leakage exceeds 212 F flashes to vapor. This flash fraction, FF, that the fraction of total should be determined using a constant iodine in the liquid that enthalpy, h, process, based on the maximum becomes airborne is equal time-dependent temperature of the sump to the fraction of the water circulating outside the containment: leakage that flashes to vapor. This flash fraction, h f1 h f2 FF, is determined FF = assuming a constant h fg enthalpy, h, process, and Where: hf1 is the enthalpy of liquid at system is based on the maximum design temperature and pressure; hf2 is the time-dependent sump enthalpy of liquid at saturation conditions water temperature.

(14.7 psia, 212ºF); and hfg is the heat of vaporization at 212ºF.

A4-36 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis A-5.5 If the temperature of the leakage is less than Conforms - Since the 212°F or the calculated flash fraction is less calculated flashing fraction than 10%, the amount of iodine that becomes is less than 10%, and airborne should be assumed to be 10% of the without a basis for total iodine activity in the leaked fluid, unless justifying a smaller value, a smaller amount can be justified based on 10% of the iodine in the the actual sump pH history and area ECCS leakage is ventilation rates. assumed to be released.

A-5.6 The radioiodine that is postulated to be Conforms - The available for release to the environment is radioiodine that is assumed to be 97% elemental and 3% postulated to be available organic. Reduction in release activity by for release to the dilution or holdup within buildings, or by ESF environment is modeled ventilation filtration systems, may be credited as 97% elemental and 3%

where applicable. Filter systems used in these organic.

applications should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

A-6 For BWRs, the main steam isolation valves Not Applicable. VEGP is a (MSIVs) have design leakage that may result PWR.

in a radioactivity release. The radiological consequences from postulated MSIV leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage.

A-7 The radiological consequences from post- Conforms - VEGP uses LOCA primary containment purging as a hydrogen recombiners for combustible gas or pressure control measure post-accident hydrogen should be analyzed. If the installed control. As such, the containment purging capabilities are containment mini-purge maintained for purposes of severe accident system is assumed to not management and are not credited in any be available for design basis analysis, radiological combustible gas consequences need not be evaluated. If the management and this primary containment purging is required within pathway is assumed to 30 days of the LOCA, the results of this remain closed following a analysis should be combined with containment isolation consequences postulated for other fission signal.

product release paths to determine the total A4-37 to Enclosure Regulatory Guide 1.183 Conformance Tables Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident)

RG Section RG Position VEGP Analysis calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

Footnote This document describes statistical Conforms - The removal A-1 formulations with differing levels of rate constants selected for uncertainty. The removal rate constants use in the LOCA selected for use in design basis calculations calculation are those that should be those that will maximize the dose will maximize the dose consequences. For BWRs, the simplified consequences.

model should be used only if the release from the core is not directed through the suppression pool. Iodine removal in the suppression pool affects the iodine species assumed by the model to be present initially.

A4-38 to Enclosure Regulatory Guide 1.183 Conformance Tables Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position VEGP Analysis B-1 Acceptable assumptions regarding core inventory Conforms - See discussions and the release of radionuclides from the fuel are in Table A.

provided in Regulatory Position 3 of this guide.

B-1.1 The number of fuel rods damaged during the Conforms - The FHA is a accident should be based on a conservative single fuel assembly dropped analysis that considers the most limiting case. from within either the This analysis should consider parameters such as Containment or the Fuel the weight of the dropped heavy load or the Handling Building, and weight of a dropped fuel assembly (plus any consistent with CLB hits attached handling grapples), the height of the another assembly damaging drop, and the compression, torsion, and shear 50 rods in the target stresses on the irradiated fuel rods. Damage to assembly. The number of adjacent fuel assemblies, if applicable (e.g., fuel rods damaged is equal events over the reactor vessel), should be to one fuel assembly plus the considered. 50 damaged rods in the target assembly (264+50=314).

B-1.2 The fission product release from the breached Exception - The fission fuel is based on Regulatory Position 3.2 of this product release is equal to guide and the estimate of the number of fuel rods the gap release, with isotopic breached. All the gap activity in the damaged fractions as given in Table rods is assumed to be instantaneously released. 2.9 of PNNL-18212 Rev. 1.

Radionuclides that should be considered include This is conservative xenons, kryptons, halogens, cesiums, and compared to RG-1.183.

rubidiums. Cycle to cycle fuel load variations are accounted for with adjustments to the core source term: +10% FDM.

B-1.3 The chemical form of radioiodine released from Conforms - The chemical the fuel to the spent fuel pool should be assumed forms of radioiodine released to be 95% cesium iodide (CsI), 4.85 percent from the fuel to the spent fuel elemental iodine, and 0.15 percent organic pool is assumed to be 95%

iodide. The CsI released from the fuel is assumed cesium iodide (CsI), 4.85 to completely dissociate in the pool water. percent elemental iodine, Because of the low pH of the pool water, the and 0.15 percent organic iodine re-evolves as elemental iodine. This is iodide. The CsI released assumed to occur instantaneously. The NRC staff from the fuel completely will consider, on a case-by-case basis, justifiable dissociates in the pool water mechanistic treatment of the iodine release from and re-evolves as elemental the pool. iodine. The dissociation and re-evolution occurs instantaneously.

B-2 If the depth of water above the damaged fuel is Conforms - Water level is 23 feet or greater, the decontamination factors for greater than 23 feet.

the elemental and organic species are 500 and 1, Therefore, the pool water is A4-39 to Enclosure Regulatory Guide 1.183 Conformance Tables Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position VEGP Analysis respectively, giving an overall effective assumed to have an overall decontamination factor of 200 (i.e., 99.5% of the decontamination factor of total iodine released from the damaged rods is 200 for the iodine isotopes retained by the water). This difference in released from the gap.

decontamination factors for elemental (99.85%)

and organic iodine (0.15%) species results in the iodine above the water being composed of 57%

elemental and 43% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1).

B-3 The retention of noble gases in the water in the Conforms - Noble gases are fuel pool or reactor cavity is negligible (i.e., not scrubbed by the pool decontamination factor of 1). Particulate water (decontamination radionuclides are assumed to be retained by the factor of 1). Particulate water in the fuel pool or reactor cavity (i.e., infinite releases are assumed to be decontamination factor). entirely scrubbed (infinite decontamination factor).

B-4.1 The radioactive material that escapes from the Conforms - For releases in fuel pool to the fuel building is assumed to be the Fuel Handling Building, released to the environment over a 2-hour time the VEGP fuel handling period. analysis considers a release to the environment over a 2-hour time period.

B-4.2 A reduction in the amount of radioactive material Conforms - no filtration is released from the fuel pool by engineered safety modeled in the FHA feature (ESF) filter systems may be taken into analysis.

account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system1 should be determined and accounted for in the radioactivity release analyses.

B-4.3 The radioactivity release from the fuel pool should Conforms - No credit for be assumed to be drawn into the ESF filtration mixing or dilution is modeled.

system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

A4-40 to Enclosure Regulatory Guide 1.183 Conformance Tables Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position VEGP Analysis B-5.1 If the containment is isolated during fuel handling Not Applicable -

operations, no radiological consequences need to Containment is not assumed be analyzed. to be isolated during fuel handling operations.

B-5.2 If the containment is open during fuel handling Not Applicable -A 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months operations, but designed to automatically isolate release from the Fuel in the event of a fuel handling accident, the Building bounds a FHA in release duration should be based on delays in containment assuming all radiation detection and completion of containment airlocks are open.

isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

B-5.3 If the containment is open during fuel handling Not Applicable - The FHA operations (e.g., personnel air lock or equipment radiological release is over a hatch is open), the radioactive material that two-hour period, but the escapes from the reactor cavity pool to the release from the Fuel containment is released to the environment over Building is closer than a a 2-hour time period. release from the containment and thus the dispersion factors for the Fuel Building release bound that for containment.

B-5.4 A reduction in the amount of radioactive material Not Applicable - No credit is released from the containment by ESF filter taken for ESF filter systems systems may be taken into account provided that to mitigate radioactive these systems meet the guidance of Regulatory material release from the Guide 1.52 and Generic Letter 99-02 (Refs. B-2 Containment.

and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

B-5.5 Credit for dilution or mixing of the activity released Not Applicable - see from the reactor cavity by natural or forced response to B-5.4 convection inside the containment may be considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

A4-41 to Enclosure Regulatory Guide 1.183 Conformance Tables Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position VEGP Analysis Footnote These analyses should consider the time for the Conforms - The FHA B-1 radioactivity concentration to reach levels calculation assumes a 600 s corresponding to the monitor setpoint, instrument delay for concentration levels line sampling time, detector response time, to reach the CREFS isolation diversion damper alignment time, and filter setpoint, an additional 8 s system actuation, as applicable. delay for CREFS components to close and an additional 90 seconds for pressurization to occur. This results in a total delay time of 698 s.

Footnote Containment isolation does not imply containment Not Applicable -

B-2 integrity as defined by technical specifications for Containment is not assumed non-shutdown modes. The term isolation is used to be isolated during fuel here collectively to encompass both containment handling operations.

integrity and containment closure, typically in place during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical specifications.

Footnote The staff will generally require that technical Not Applicable - No credit is B-3 specifications allowing such operations include taken for containment administrative controls to close the airlock, hatch, isolation.

or open penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur.

Radiological analyses should generally not credit this manual isolation.

A4-42 to Enclosure Regulatory Guide 1.183 Conformance Tables Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident)

RG Section RG Position VEGP Analysis E-1 Assumptions acceptable to the NRC staff Conforms - See regarding core inventory and the release of discussions in Table A.

radionuclides from the fuel are provided in Regulatory Position 3 of this regulatory guide. The 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. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

E-2 If no or minimal fuel damage is postulated for the Consistent with the VEGP limiting event, the activity released should be the current licensing basis a maximum coolant activity allowed by the technical leaking fuel term is specifications. Two cases of iodine spiking should conservatively included with be assumed. the two cases of iodine spiking.

E-2.1 A reactor transient has occurred prior to the Conforms - The Main postulated main steam line break (MSLB) and has Steam Line Break Accident raised the primary coolant iodine concentration to dose calculation includes a the maximum value (typically 60 µCi/gm DE I-131) case for a preaccident permitted by the technical specifications (i.e., a iodine spike with the preaccident iodine spike case). maximum iodine concentration permitted by the VEGP technical specifications.

E-2.2 The primary system transient associated with the Conforms - The Main MSLB causes an iodine spike in the primary Steam Line Break Accident system. The increase in primary coolant iodine dose calculation includes a concentration is estimated using a spiking model case for a concurrent iodine that assumes that the iodine release rate from the spike causing the iodine fuel rods to the primary coolant (expressed in release rate from the fuel curies per unit time) increases to a value 500 times rods to the RCS to increase greater than the release rate corresponding to the to a value 500 times greater iodine concentration at the equilibrium value than the release rate that (typically 1.0 µCi/gm DE I-131) specified in yields the maximum technical specifications (i.e., concurrent iodine equilibrium iodine spike case). A concurrent iodine spike need not be concentration specified in considered if fuel damage is postulated. The the technical specifications.

assumed iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The iodine spike duration is Shorter spike durations may be considered on a 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins.

A4-43 to Enclosure Regulatory Guide 1.183 Conformance Tables Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident)

RG Section RG Position VEGP Analysis E-3 The activity released from the fuel should be Conforms - The initial assumed to be released instantaneously and activity from the fuel is homogeneously through the primary coolant. assumed to be released instantaneously and homogeneously to the reactor coolant system.

E-4 The chemical form of radioiodine released from the Conforms - The iodine fuel should be assumed to be 95% cesium iodide releases from the steam (CsI), 4.85 percent elemental iodine, and 0.15 generators to the percent organic iodide. Iodine releases from the environment are 97%

steam generators to the environment should be elemental and 3% organic assumed to be 97% elemental and 3% organic. for the pre-accident case These fractions apply to iodine released as a result and the concurrent iodine of fuel damage and to iodine released during spike case.

normal operations, including iodine spiking.

E-5.1 For facilities that have not implemented alternative Conforms -The assumed repair criteria (see Ref. E-1, DG-1074), the primary-to-secondary leak primary-to-secondary leak rate in the steam rate in the three intact generators should be assumed to be the leak rate steam generators are 0.65 limiting condition for operation specified in the gpm (936 gallons per day).

technical specifications. For facilities with This is conservative relative traditional generator specifications (both per to VEGP TS 3.4.13 which generator and total of all generators), the leakage allows 150 gallons per day should be apportioned between affected and per Steam Generator.

unaffected steam generators in such a manner that the calculated dose is maximized.

E-5.2 The density used in converting volumetric leak Conforms - The assumed rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) density is 62.4 lbm/ft3.

should be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

E-5.3 The primary-to-secondary leakage should be Conforms - For the faulted assumed to continue until the primary system steam generator, primary-pressure is less than the secondary system to-secondary leakage pressure, or until the temperature of the leakage is continues for the duration of less than 100°C (212°F). The release of the event. The release from radioactivity from unaffected steam generators the unaffected steam should be assumed to continue until shutdown generators continues until cooling is in operation and releases from the steam the RHR system is placed generators have been terminated. in service in 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

A4-44 to Enclosure Regulatory Guide 1.183 Conformance Tables Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident)

RG Section RG Position VEGP Analysis E-5.4 All noble gas radionuclides released from the Conforms - All noble gases primary system are assumed to be released to the are released from the steam environment without reduction or mitigation. generator water without credit for scrubbing.

E-5.5 The transport model described in this section Conforms - See below.

should be utilized for iodine and particulate releases from the steam generators. This model is shown in Figure E-1 and summarized below:

E-5.5.1 A portion of the primary-to-secondary leakage will Conforms - The leakage of flash to vapor, based on the thermodynamic the faulted steam generator conditions in the reactor and secondary coolant. is modeled as a direct vapor

During periods of steam generator dryout, flow from the RCS to the all of the primary-to-secondary leakage is environment without assumed to flash to vapor and be released partitioning. For the intact to the environment with no mitigation. steam generators, primary-

With regard to the unaffected steam to-secondary leakage mixes generators used for plant cooldown, the with the secondary water primary-to-secondary leakage can be without flashing for the assumed to mix with the secondary water duration of the event.

without flashing during periods of total tube submergence.

E-5.5.2 The leakage that immediately flashes to vapor will Conforms - For rise through the bulk water of the steam generator conservatism, no credit is and enter the steam space. Credit may be taken taken for scrubbing.

for scrubbing in the generator, using the models in NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident (Ref. E-2), during periods of total submergence of the tubes.

E-5.5.3 The leakage that does not immediately flash is Conforms - The leakage assumed to mix with the bulk water. that does not immediately flash mixes with the bulk water.

E-5.5.4 The radioactivity in the bulk water is assumed to Conforms - For flows out of become vapor at a rate that is the function of the the intact SGs, radioactivity steaming rate and the partition coefficient. A to the environment is a partition coefficient for iodine of 100 may be function of the steaming assumed. The retention of particulate rate, and the iodine partition radionuclides in the steam generators is limited by factor is assumed to be the moisture carryover from the steam generators. 100. Moisture carryover is modeled at 0.32%.

E-5.6 Operating experience and analyses have shown Conforms - The steam that for some steam generator designs, tube generator with the faulted uncovery may occur for a short period following main steamline in the MSLB any reactor trip (Ref. E-3). The potential impact of accident is assumed to blow tube uncovery on the transport model parameters completely dry, causing a A4-45 to Enclosure Regulatory Guide 1.183 Conformance Tables Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident)

RG Section RG Position VEGP Analysis (e.g., flash fraction, scrubbing credit) needs to be direct release of considered. The impact of emergency operating radioactivity from that procedure restoration strategies on steam source to the environment.

generator water levels should be evaluated.

Footnote Facilities licensed with, or applying for, alternative Conforms - VEGP is E-1 repair criteria (ARC) should use this section in licensed to ARC.

conjunction with the guidance that is being developed in Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity, for acceptable assumptions and methodologies for performing radiological analyses.

Footnote The activity assumed in the analysis should be Consistent with the VEGP E-2 based on the activity associated with the projected current licensing basis the fuel damage or the maximum technical maximum RCS specification values, whichever maximizes the concentrations allowed by radiological consequences. In determining dose TS are used.

equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

A4-46 to Enclosure Regulatory Guide 1.183 Conformance Tables Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident)

RG Section RG Position VEGP Analysis F-1 Assumptions acceptable to the NRC staff Conforms - See discussions regarding core inventory and the release of in Table A.

radionuclides from the fuel are in Regulatory Position 3 of this guide. The 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.

F-2 If no or minimal2 fuel damage is postulated for the Conforms-the maximum limiting event, the activity released should be the RCS concentrations allowed maximum coolant activity allowed by technical by TS are used.

specification. Two cases of iodine spiking should be assumed.

F-2.1 A reactor transient has occurred prior to the Conforms - Case 1 is a pre-postulated steam generator tube rupture (SGTR) accident spike using the and has raised the primary coolant iodine maximum Dose Equivalent concentration to the maximum value (typically 60 Iodine permitted by the

µCi/gm DE I-131) permitted by the technical VEGP Technical specifications (i.e., a preaccident iodine spike Specifications.

case).

F-2.2 The primary system transient associated with the Conforms - The concurrent SGTR causes an iodine spike in the primary iodine spike case assumes system. The increase in primary coolant iodine the RCS transient concentration is estimated using a spiking model associated with the accident that assumes that the iodine release rate from the creates an iodine spike, fuel rods to the primary coolant (expressed in causing the iodine release curies per unit time) increases to a value 335 rate from the fuel rods to the times greater than the release rate corresponding RCS to increase to a value to the iodine concentration at the equilibrium 335 times greater than the value (typically 1.0 µCi/gm DE I-131) specified in release rate that yields the technical specifications (i.e., concurrent iodine maximum allowable spike case). A concurrent iodine spike need not equilibrium iodine be considered if fuel damage is postulated. The concentration specified in assumed iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . the technical specifications.

Shorter spike durations may be considered on a A 20-hour release duration is case-by-case basis if it can be shown that the modeled.

activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins.

F-3 The activity released from the fuel, if any, should Conforms - Mixing in the be assumed to be released instantaneously and primary coolant is assumed homogeneously through the primary coolant. to be instantaneously and homogeneously.

F-4 Iodine releases from the steam generators to the Conforms - The iodine environment should be assumed to be 97% released to the environment elemental and 3% organic. is assumed to be 97%

elemental and 3% organic.

A4-47 to Enclosure Regulatory Guide 1.183 Conformance Tables Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident)

RG Section RG Position VEGP Analysis F-5.1 The primary-to-secondary leak rate in the steam Conforms - The assumed generators should be assumed to be the leak rate primary-to-secondary leak limiting condition for operation specified in the rate in the three intact steam technical specifications. The leakage should be generators are 0.7 gpm apportioned between affected and unaffected (1008 gallons per day). This steam generators in such a manner that the is conservative relative to calculated dose is maximized. VEGP TS 3.4.13 which allows 150 gallons per day per Steam Generator.

F-5.2 The density used in converting volumetric leak Conforms - The assumed rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) density is 62.4 lbm/ft3.

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

F-5.3 The primary-to-secondary leakage should be Conforms - It is assumed assumed to continue until the primary system that the RHR system is pressure is less than the secondary system placed in service at 20 pressure, or until the temperature of the leakage hours, terminating the is less than 100°C (212°F). The release of accident.

radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

F-5.4 The release of fission products from the Conforms - The SGTR secondary system should be evaluated with the assumes a concurrent LOOP assumption of a coincident loss of offsite power. to maximize the release to the environment.

F-5.5 All noble gas radionuclides released from the Conforms - Noble gases are primary system are assumed to be released to modeled as going directly to the environment without reduction or mitigation. the environment without reduction or mitigation.

F-5.6 The transport model described in Regulatory Conforms - consistent with Positions 5.5 and 5.6 of Appendix E should be the CLB, flashing fraction is utilized for iodine and particulates. as calculated by the LOFTRAN code, and particulate transport is based on the maximum moisture carryover.

Footnote Facilities licensed with, or applying for, alternative Conforms - VEGP is F-1 repair criteria (ARC) should use this section in licensed to ARC.

conjunction with the guidance that is being developed in Draft Regulatory Guide DG-1074, A4-48 to Enclosure Regulatory Guide 1.183 Conformance Tables Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident)

RG Section RG Position VEGP Analysis Steam Generator Tube Integrity (USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

Footnote The activity assumed in the analysis should be Conforms-The initial RCS F-2 based on the activity associated with the activities are based on the projected fuel damage or the maximum technical maximum allowable values specification values, whichever maximizes the by TS.

radiological consequences. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

A4-49 to Enclosure Regulatory Guide 1.183 Conformance Tables Table F: Conformance With Regulatory Guide 1.183 Appendix G (Locked Rotor Accident)

RG Section RG Position VEGP Analysis G-1 Assumptions acceptable to the NRC staff Conforms - See discussions regarding core inventory and the release of in Table A.

radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The 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.

G-2 If no fuel damage is postulated for the limiting Conforms - The transient event, a radiological analysis is not required as causes fuel damage and so the consequences of this event are bounded by a radiological analysis is the consequences projected for the main steam provided.

line break outside containment.

G-3 The activity released from the fuel should be Conforms - The gap activity assumed to be released instantaneously and in the damaged rods is homogeneously through the primary coolant. instantaneously released to and uniformly mixed within the reactor coolant system at the onset of the accident.

G-4 The chemical form of radioiodine released from Conforms - The iodine the fuel should be assumed to be 95% cesium releases from the steam iodide (CsI), 4.85 percent elemental iodine, and generators to the 0.15 percent organic iodide. Iodine releases from environment are 97%

the steam generators to the environment should elemental and 3% organic be assumed to be 97% elemental and 3% for the pre-accident case organic. These fractions apply to iodine released and the concurrent iodine as a result of fuel damage and to iodine released spike case, including during normal operations, including iodine spiking. damaged fuel.

G-5.1 The primary-to-secondary leak rate in the steam Conforms - Leakage is 1 generators should be assumed to be the leak- gpm, which is bounding over rate-limiting condition for operation specified in the the Technical Specification technical specifications. The leakage should be limit of 150 gallons per day apportioned between the steam generators in per steam generator.

such a manner that the calculated dose is maximized.

G-5.2 The density used in converting volumetric leak Conforms - The assumed rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) density is 62.4 lbm/ft3.

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

G-5.3 The primary-to-secondary leakage should be Conforms - The accident assumed to continue until the primary system terminates after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months pressure is less than the secondary system and the RHR system is pressure, or until the temperature of the leakage placed in service.

A4-50 to Enclosure Regulatory Guide 1.183 Conformance Tables Table F: Conformance With Regulatory Guide 1.183 Appendix G (Locked Rotor Accident)

RG Section RG Position VEGP Analysis is less than 100°C (212°F). The release of radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

G-5.4 The release of fission products from the Conforms - The Locked secondary system should be evaluated with the Rotor Accident assumes a assumption of a coincident loss of offsite power. concurrent LOOP to maximize the release to the environment.

G-5.5 All noble gas radionuclides released from the Conforms - Noble gases are primary system are assumed to be released to the assumed to leak directly to environment without reduction or mitigation. the environment without holdup in the SG.

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

Footnote Facilities licensed with, or applying for, alternative Conforms - VEGP is G-1 repair criteria (ARC) should use this section in licensed to ARC.

conjunction with the guidance that is being developed in Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity (USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

A4-51 to Enclosure Regulatory Guide 1.183 Conformance Tables 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 Position 3 of this guide. For the rod ejection based upon Appendix H, the amount accident, the release from the breached fuel is of damaged fuel, and the assumption based on the estimate of the number of fuel that 10% of the core inventory of rods breached and the assumption that 10% noble gases and iodine isotopes are of the core inventory of the noble gases and in the fuel rod gap. Alkali metals iodines is in the fuel gap. The release (12%) and other halogens (5%) are attributed to fuel melting is based on the also assumed to be in the fuel gap, fraction of the fuel that reaches or exceeds consistent with Table 3.

the initiation temperature for fuel melting and the assumption that 100% of the noble gases For releases from containment and 25% of the iodines contained in that involving fuel melting, 100% of the fraction are available for release from noble gases and 25% of the iodine containment. For the secondary system isotopes contained in the portion of release pathway, 100% of the noble gases the fuel that melts is available for and 50% of the iodines in that fraction are release from containment. For released to the reactor coolant. 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 event, a radiological analysis is not required is postulated for this event.

as 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.

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

A4-52 to Enclosure Regulatory Guide 1.183 Conformance Tables H-4 The chemical form of radioiodine released to Conforms - The chemical form of the containment atmosphere should be radioiodine released to the assumed to be 95% cesium iodide (CsI), containment atmosphere is assumed 4.85% elemental iodine, and 0.15% organic to be 95% cesium iodide, 4.85%

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

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.

A4-53 to Enclosure Loss-of-Coolant Accident Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 5 Loss-of-Coolant Accident Analysis A5-1 to Enclosure Loss-of-Coolant Accident Analysis LOSS-OF-COOLANT ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.15, Version 2 Method/Computer Program Used: RADTRAD Version 3.10 Regulatory Guidance: RG-1.183, including Appendix A Model Discussion The calculation was performed in four parts, evaluating the contributions from four separate release paths: Containment mini-purge, Containment Leakage, ECCS Leakage Outside of Containment, and potential leakage from the Refueling Water Storage Tank (RWST). The dose contributions from each of these pathways were summed to obtain the doses to the Main Control Room (MCR), the Exclusion Area Boundary (EAB), and the Low Population Zone (LPZ).

The accident duration is 30 days, per VEGP Current Licensing Basis (CLB).

Results and Acceptance Limits EAB LPZ Control Room Release (rem TEDE), Max. 2 hr (rem TEDE) (rem TEDE)

Containment Purge 0.001 0.001 0.001 Containment Leakage 6.6 5.1 2.2 ECCS Leakage 1.6 3.8 1.6 RWST Back-leakage 0.18 0.69 0.36 MCR External Sources n/a n/a 0.3 Total 8.4 9.6 4.4 Acceptance Limit 25 25 5 (Note that rounding is applied to all values)

Key Assumptions and Inputs Source Term Parameters Parameter Value Core Power Level: 3636 MWt (includes uncertainty)

Initial Core Source Term Includes 10% Fuel Design Margin (FDM) added to the nominal activity A5-2 to Enclosure Loss-of-Coolant Accident Analysis Table 1 - Core Source Term Core Core Source Source Nuclide Nuclide Term Term (Ci) (Ci)

Kr-83m 1.29E+07 Rh-103m 1.69E+08 Kr-85 1.12E+06 Rh-105 1.07E+08 Kr-85m 2.74E+07 Rh-106 6.16E+07 Kr-87 5.40E+07 Ru-103 1.69E+08 Kr-88 7.23E+07 Ru-105 1.18E+08 Xe-131m 1.42E+06 Ru-106 5.39E+07 Xe-133 2.15E+08 Tc-99 1.48E+03 Xe-133m 6.83E+06 Tc-99m 1.77E+08 Xe-135 5.10E+07 Ce-141 1.79E+08 Xe-135m 4.59E+07 Ce-143 1.65E+08 Xe-138 1.87E+08 Ce-144 1.34E+08 Br-82 3.41E+05 Np-237 3.44E+01 Br-83 1.28E+07 Np-238 4.27E+07 Br-84 2.32E+07 Np-239 2.07E+09 I-130 2.04E+06 Pu-238 3.02E+05 I-131 1.07E+08 Pu-239 3.09E+04 I-132 1.57E+08 Pu-240 4.50E+04 I-133 2.20E+08 Pu-241 1.30E+07 I-134 2.47E+08 Pu-242 2.06E+02 I-135 2.10E+08 Pu-243 4.22E+07 Cs-134 1.67E+07 Am-241 1.28E+04 Cs-134m 4.50E+06 Am-242 7.06E+06 Cs-135 4.51E+01 Am-243 2.53E+03 Cs-136 5.37E+06 Cm-242 3.69E+06 Cs-137 1.18E+07 Cm-244 3.71E+05 A5-3 to Enclosure Loss-of-Coolant Accident Analysis Core Core Source Source Nuclide Nuclide Term Term (Ci) (Ci)

Cs-138 2.04E+08 Eu-154 6.80E+05 Rb-86 2.07E+05 Eu-155 2.87E+05 Rb-88 7.36E+07 Eu-156 2.46E+07 Rb-89 9.64E+07 La-140 1.96E+08 Sb-124 7.92E+04 La-141 1.76E+08 Sb-125 8.59E+05 La-142 1.70E+08 Sb-126 5.11E+04 La-143 1.63E+08 Sb-127 9.75E+06 Nb-95 1.84E+08 Sb-129 3.03E+07 Nb-95m 2.09E+06 Te-125m 1.82E+05 Nb-97 1.83E+08 Te-127 9.55E+06 Nb-97m 1.73E+08 Te-127m 1.58E+06 Nd-147 7.08E+07 Te-129 2.84E+07 Pm-147 1.77E+07 Te-129m 5.44E+06 Pm-148 1.93E+07 Te-131 9.10E+07 Pm-148m 4.01E+06 Te-131m 2.07E+07 Pm-149 6.53E+07 Te-132 1.53E+08 Pm-151 2.06E+07 Te-133 1.17E+08 Pr-143 1.61E+08 Te-133m 1.03E+08 Pr-144 1.35E+08 Te-134 1.96E+08 Pr-144m 1.88E+06 Sr-89 1.02E+08 Sm-151 4.47E+04 Sr-90 8.70E+06 Sm-153 4.93E+07 Sr-91 1.27E+08 Y-90 9.08E+06 Sr-92 1.37E+08 Y-91 1.34E+08 Ba-137m 1.12E+07 Y-91m 7.39E+07 Ba-139 1.95E+08 Y-92 1.38E+08 Ba-140 1.89E+08 Y-93 1.57E+08 A5-4 to Enclosure Loss-of-Coolant Accident Analysis Core Core Source Source Nuclide Nuclide Term Term (Ci) (Ci)

Ba-141 1.75E+08 Y-95 1.72E+08 Mo-99 2.00E+08 Zr-95 1.82E+08 Pd-109 3.49E+07 Zr-97 1.82E+08 Parameter Value Initial RCS Source Term Noble Gases, Halogens, and Alkali Metals I-131 through I-135 1 µCi/g Dose Equivalent I-133 Noble Gases (except Kr-83m) 280 µCi/g Dose Equivalent Xe-133 consistent with TSTF-490 Rev. 0 Other Nuclides Based on 1% Fuel Clad Defects; includes 10% Fuel Design Margin RCS Mass 2.258E+08 grams Release Fractions Per RG-1.183 Release Timing Per RG-1.183 A5-5 to Enclosure Loss-of-Coolant Accident Analysis Table 2 - RCS Source Term RCS Source RCS Source Nuclide Term Nuclide Term

(µCi/g) (µCi/g)

Kr-83m 4.69E-01 I-131 7.61E-01 Kr-85 3.84E+00 I-132 7.79E-01 Kr-85m 7.46E-01 I-133 1.16E+00 Kr-87 4.87E-01 I-134 1.66E-01 Kr-88 1.64E+00 I-135 6.40E-01 Xe-131m 1.46E+00 Cs-134 1.99E+00 Xe-133 1.15E+02 Cs-134m 2.46E-02 Xe-133m 1.59E+00 Cs-135 0.00E+00 Xe-135 3.47E+00 Cs-136 2.72E+00 Xe-135m 2.19E-01 Cs-137 1.92E+00 Xe-138 2.77E-01 Cs-138 1.07E+00 Br-82 5.12E-03 Rb-86 2.68E-02 Br-83 9.88E-02 Rb-88 4.33E+00 Br-84 4.66E-02 Rb-89 1.01E-01 I-130 2.19E-02 Note 1: Cs-135 activity omitted as it is considered to be a negligible contributor to doses with respect to the other included nuclides.

Containment Leakage Parameters Parameter Value Containment Volume 2.930E+06 cubic feet Sprayed Volume 2.30E+06 cubic feet Unsprayed Volume 6.30E+05 cubic feet Containment Leakage 0.21% of volume per day for first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Tech. Spec. Leakage plus 5% margin) 0.105% of volume per day for remainder Containment Leakage Filtration None Containment Long Term Sump pH pH 7.0 (no re-evolution of Iodine)

Containment spray removal , Elemental 13.7 hr-1 Containment spray removal , Aerosol 5.34 hr-1 Containment Spray Organic removal None A5-6 to Enclosure Loss-of-Coolant Accident Analysis Natural Deposition, Aerosol only 0.1 hr-1 after sprays are terminated Containment Spray Start 10 seconds Containment Spray Stop 2 Hours Containment Spray Flow 2,500 gal/min Iodine Chemical Form 95% Cesium Iodide, 4.85% elemental, 0.15% organic Containment Purge Leakage Parameter Value Iodine Chemical Form 95% Cesium Iodide, 4.85% elemental, 0.15% organic Containment Purge Filtration None Removal by wall deposition 0%

Removal by Sprays 0%

Containment Purge Isolation 30 seconds Containment Purge Flowrate 5000 CFM ECCS Leakage Parameter Value Sump Volume 114,922 cubic feet Sump temperature Varies, max is 260 ºF ECCS Leakage Initiation Time 30 minutes ECCS Leakage Iodine Flashing Factor 10%

Iodine Species ECCS Leakage Released to the Atmosphere Elemental 97%

Organic 3%

ECCS Leakage Rate 2.0 gal/min RWST Leakage Parameters Parameter Value ECCS Recirculation Start Time 30 minutes Iodine Species ECCS Leakage Released to the Atmosphere from the RWST Elemental 100%

Organic 0%

ECCS Leakage Rate to the RWST 7 gal/min RWST Leakage Iodine Flashing Factors Varies with temperature and pH between 0% and 10.03%

RWST Capacity 731,000 gallons RWST Volume at Transfer to Recirculation 95,837.7 gallons A5-7 to Enclosure Loss-of-Coolant Accident Analysis CR Parameters:

Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 11.3 Seconds CR Pressurization Mode Initiation Automatic at 99.3 Seconds CR Ventilation System Normal Flow Rate 2,575 cfm < 11.3 seconds CR Ventilation System Filtered Makeup Rate 1,800 cfm > 99.3 seconds CR Ventilation System Recirculation Flow Rate 31,000 cfm > 99.3 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 > 11.3 seconds CR Ingress/Egress Unfiltered In-leakage 10 cfm > 0 seconds CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Atmospheric Dispersion Factors (sec/m3):

All Release Pathways Time (hr) EAB* LPZ CR 0-2 1.8E-4 7.20E-5 2.22E-03 2-8 - 3.30E-5 1.55E-03 8 - 24 - 2.20E-5 6.57E-04 24 - 96 - 9.20E-6 5.80E-04 96 - 720 - 2.70E-6 4.47E-04

Dispersion Factor applied for entire duration of transient in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec 24 Hours - 30 Days 2.3E-04 m3/sec A5-8 to Enclosure Fuel Handling Accident Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 6 Fuel Handling Accident Analysis A6-1 to Enclosure Fuel Handling Accident Analysis FUEL HANDLING ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.16, Version 3 Method/Computer Program Used: RADTRAD Version 3.10 Regulatory Guidance: RG-1.183, including Appendix B Model Discussion The calculation was performed to address a fuel handling accident (FHA) in the SFP area of the Auxiliary Building. For a FHA in containment, the X/Qs comparisons between an accident in the fuel building and for one in containment shows the FHA in the fuel building is more limiting. As a result, only an accident in the fuel building is analyzed. Doses in the CR are accumulated over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Releases from the damaged fuel are completed in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The activity is released to the environment through the plant vent stack at the elevation of the fuel building roof to maximize the releases to the CR. No credit is taken for filtration of the iodine isotopes is assumed. By inspection, doses from this accident bound the doses from an accident in containment.

Results and Acceptance Limits EAB LPZ Control Room Release (rem TEDE) (rem TEDE) (rem TEDE)

Spent Fuel Pool 1.0 0.4 3.9 Acceptance Limit 6.3 6.3 5 (Note that rounding is applied to all values)

Key Assumptions and Inputs Source Term Parameters Parameter Value Reactor Power Level 3636 MWt (includes uncertainty)

Radial Peaking Factor 1.7 Fuel Movement Time 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months post shutdown.

Number of Fuel Assemblies 193 Number of Fuel Rods/Assembly 264 Number of Rods in Core 50,952 Number of Damaged Assemblies 1 + 50 rods in target assembly Number of Damaged Fuel Rods 314 Fuel Handling Building Volume 1 cu ft Fuel Building Exhaust Rate 0.115 cfm (ensures all activity released in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months )

A6-2 to Enclosure Fuel Handling Accident Analysis Table 1 - Source Term 70-Hour Core Released Activity Gap Nuclide Source Term DF (314 Rods)

Fraction (Ci) (Ci)

Kr-83m 9.73E-02 8.00E-02 1 8.23E-05 Kr-85 1.12E+06 3.80E-01 1 4.51E+03 Kr-85m 5.49E+02 3.80E-01 1 2.20E+00 Kr-87 1.48E-09 8.00E-02 1 1.25E-12 Kr-88 2.75E+00 8.00E-02 1 2.33E-03 Xe-131m 1.36E+06 8.00E-02 1 1.15E+03 Xe-133 1.73E+08 8.00E-02 1 1.47E+05 Xe-133m 3.99E+06 8.00E-02 1 3.38E+03 Xe-135 2.59E+06 8.00E-02 1 2.19E+03 Xe-135m 2.13E+04 8.00E-02 1 1.80E+01 Xe-138 0.00E+00 8.00E-02 1 0.00E+00 Br-82 8.65E+04 5.00E-02 200 2.28E-01 Br-83 2.32E-02 5.00E-02 200 6.12E-08 Br-84 0.00E+00 5.00E-02 200 0.00E+00 I-130 4.04E+04 5.00E-02 200 1.07E-01 I-131 8.59E+07 8.00E-02 200 3.63E+02 I-132 8.48E+07 9.00E-02 200 4.03E+02 I-133 2.19E+07 5.00E-02 200 5.79E+01 I-134 0.00E+00 5.00E-02 200 0.00E+00 I-135 1.30E+05 5.00E-02 200 3.44E-01 Released Activity = (Radial Peaking Factor) * [(70-Hour Core Source Term) * (Gap Fraction) *

(Number of Damaged Rods)/(Number of Rods in Core)]/(Net Decontamination Factor)

Overlaying Pool Depth 23 feet Iodine Chemical Form 0% Aerosol, 99.85% Elemental 0.15% Organic Net Decontamination Factor 200 Fuel Design Margin 10%

While it is generally considered that Br in the Halogen group is in particulate form (RG-1.183 Section 3.5) and would be entrained in the 23 ft pool water depth, it is conservatively considered here. The conservative treatment of Br is considered to alleviate any considerations due to interpretation of PNNL-18212 Rev. 1 Table 2.9 for the Other Halogens.

A6-3 to Enclosure Fuel Handling Accident Analysis CR Parameters:

Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 0.169 hour0.00196 days 0.0469 hours 2.794312e-4 weeks 6.43045e-5 months CR Pressurization Mode Initiation Automatic at 0.194 hour0.00225 days 0.0539 hours 3.207672e-4 weeks 7.3817e-5 months CR Ventilation System Normal Flow Rate 2,575 cfm < 0.169 hour0.00196 days 0.0469 hours 2.794312e-4 weeks 6.43045e-5 months CR Ventilation System Filtered Makeup Rate 1,800 cfm > 0.194 hour0.00225 days 0.0539 hours 3.207672e-4 weeks 7.3817e-5 months CR Ventilation System Recirculation Flow Rate 31,000 cfm > 0.194 hour0.00225 days 0.0539 hours 3.207672e-4 weeks 7.3817e-5 months CR Unfiltered In-leakage 180 cfm > 0.169 hour0.00196 days 0.0469 hours 2.794312e-4 weeks 6.43045e-5 months CR Ingress/Egress Unfiltered In-leakage 10 cfm > 0 seconds CR Ventilation System Charcoal Filter Efficiencies (Supply and Recirculation use the same filter)

All Iodine Species 99%

Particulates 99%

CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 Atmospheric Dispersion Factors Time (hr) X/Q (sec/m3)

EAB* LPZ MCR 0-2 1.80E-04 7.20E-05 6.01E-03 2-8 - 3.30E-05 4.44E-03 8 - 24 - 2.2E-05 1.71E-03

Applied for entire duration of event in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec A6-4 to Enclosure Main Steam Line Break Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 7 Main Steam Line Break Analysis A7-1 to Enclosure Main Steam Line Break Analysis MAIN STEAM LINE BREAK ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.17, Version 2 Method/Computer Program Used: RADTRAD Version 3.10 Regulatory Guidance: RG-1.183, including Appendix E Model Discussion:

The calculation was performed to address a Main Steam Line Break (MSLB). Per RG-1.183, two cases are considered for the dose-equivalent I-131 (DEI) concentrations in the Reactor Coolant System (RCS):

1. Pre-Accident Iodine Spike - a reactor transient occurs prior to the accident and raises the RCS iodine concentration to the maximum value permitted by the technical specifications.

2. Concurrent Iodine Spike - the RCS transient associated with the accident creates an iodine spike, causing the iodine release rate from the fuel rods to the RCS to increase to a value 500 times greater than the release rate that yields the equilibrium iodine concentration specified in the technical specifications.

Primary to Secondary leakage (consistent with CLB) is assumed to be 0.35 gallons per minute (gpm) to the faulted steam generator (SG), and 0.65 gpm (total) going to the three intact SGs.

It is postulated that the MSLB causes the associated faulted SG to blow dry, releasing activity directly to the environment through the broken main steam line. Activity from three intact SGs released to the environment via steaming until the primary system (RCS) is placed on RHR cooling (assumed at 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

Doses for the Pre-Accident Iodine Spike Case and the Concurrent Iodine Spike Case were calculated, with results shown below.

Results and Acceptance Limits Case Location Dose (Rem TEDE)

Calculated Limit Pre-Accident Iodine EAB <0.1 25 Spike LPZ <0.1 25 Control Room <0.1 5 Concurrent Iodine EAB 0.2 2.5 Spike LPZ 0.2 2.5 Control Room 0.2 5 (Note that rounding is applied to all values)

A7-2 to Enclosure Main Steam Line Break Analysis Key Assumptions and Inputs Source Terms Table 1 - RCS Source Terms Pre-Accident Iodine Spike Concurrent Iodine Spike RCS Source RCS Source RCS Source Term RCS Source Nuclide Term Term (Ci) (µCi/g) Term (Ci)

(µCi/g)

Kr-83m 4.69E-01 1.059E+02 4.69E-01 1.059E+02 Kr-85 3.84E+00 8.669E+02 3.84E+00 8.669E+02 Kr-85m 7.46E-01 1.684E+02 7.46E-01 1.684E+02 Kr-87 4.87E-01 1.100E+02 4.87E-01 1.100E+02 Kr-88 1.64E+00 3.696E+02 1.64E+00 3.696E+02 Xe-131m 1.46E+00 3.292E+02 1.46E+00 3.292E+02 Xe-133 1.15E+02 2.594E+04 1.15E+02 2.594E+04 Xe-133m 1.59E+00 3.597E+02 1.59E+00 3.597E+02 Xe-135 3.47E+00 7.838E+02 3.47E+00 7.838E+02 Xe-135m 2.19E-01 4.940E+01 2.19E-01 4.940E+01 Xe-138 2.77E-01 6.245E+01 2.77E-01 6.245E+01 Br-82 5.12E-03 1.157E+00 5.12E-03 1.157E+00 Br-83 9.88E-02 2.230E+01 9.88E-02 2.230E+01 Br-84 4.66E-02 1.051E+01 4.66E-02 1.051E+01 I-130 2.19E-02 4.938E+00 2.19E-02 4.938E+00 I-131 4.56E+01 1.030E+04 7.61E-01 1.717E+02 I-132 4.68E+01 1.056E+04 7.79E-01 1.760E+02 I-133 6.95E+01 1.569E+04 1.16E+00 2.615E+02 I-134 9.95E+00 2.246E+03 1.66E-01 3.744E+01 I-135 3.84E+01 8.677E+03 6.40E-01 1.446E+02 Cs-134 1.99E+00 4.491E+02 1.99E+00 4.491E+02 Cs-134m 2.46E-02 5.549E+00 2.46E-02 5.549E+00 Cs-135 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Cs-136 2.72E+00 6.135E+02 2.72E+00 6.135E+02 Cs-137 1.92E+00 4.337E+02 1.92E+00 4.337E+02 Cs-138 1.07E+00 2.409E+02 1.07E+00 2.409E+02 Rb-86 2.68E-02 6.041E+00 2.68E-02 6.041E+00 Rb-88 4.33E+00 9.766E+02 4.33E+00 9.766E+02 Rb-89 1.01E-01 2.282E+01 1.01E-01 2.282E+01 A7-3 to Enclosure Main Steam Line Break Analysis Table 2 - Concurrent Spike RCS Iodine Appearance Rate Nominal Appearance 500* Nominal Isotope Rate, Ci/hr Appearance Rate, Ci/hr I-131 1.75E+01 8.75E+03 I-132 7.05E+01 3.53E+04 I-133 3.44E+01 1.72E+04 I-134 3.33E+01 1.67E+04 I-135 2.95E+01 1.48E+04 Table 3 - Secondary Side Source Terms Intact Steam Generators Faulted Steam Generator Secondary Secondary Secondary Secondary Side Source Side Source Nuclide Side Source Side Source Term Term Term (Ci) Term (Ci)

(µCi/g) (µCi/g)

Kr-83m 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Kr-85 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Kr-85m 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Kr-87 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Kr-88 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-131m 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-133 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-133m 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-135 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-135m 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Xe-138 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Br-82 5.12E-04 6.842E-02 5.12E-04 2.281E-02 Br-83 9.88E-03 1.318E+00 9.88E-03 4.395E-01 Br-84 4.66E-03 6.215E-01 4.66E-03 2.072E-01 I-130 2.19E-03 2.919E-01 2.19E-03 9.731E-02 I-131 7.61E-02 1.015E+01 7.61E-02 3.385E+00 I-132 7.79E-02 1.040E+01 7.79E-02 3.468E+00 I-133 1.16E-01 1.546E+01 1.16E-01 5.154E+00 I-134 1.66E-02 2.214E+00 1.66E-02 7.378E-01 I-135 6.40E-02 8.550E+00 6.40E-02 2.850E+00 Cs-134 1.99E-01 2.655E+01 1.99E-01 8.850E+00 Cs-134m 2.46E-03 3.281E-01 2.46E-03 1.094E-01 Cs-135 0.00E+00 0.000E+00 0.00E+00 0.000E+00 Cs-136 2.72E-01 3.627E+01 2.72E-01 1.209E+01 Cs-137 1.92E-01 2.564E+01 1.92E-01 8.547E+00 Cs-138 1.07E-01 1.424E+01 1.07E-01 4.747E+00 Rb-86 2.68E-03 3.571E-01 2.68E-03 1.190E-01 A7-4 to Enclosure Main Steam Line Break Analysis Intact Steam Generators Faulted Steam Generator Secondary Secondary Secondary Secondary Side Source Side Source Nuclide Side Source Side Source Term Term Term (Ci) Term (Ci)

(µCi/g) (µCi/g)

Rb-88 4.33E-01 5.774E+01 4.33E-01 1.925E+01 Rb-89 1.01E-02 1.349E+00 1.01E-02 4.498E-01 Physical Parameters Parameter Value RCS Mass 2.258E8 grams RCS Volume 1.11E4 cubic feet Intact SG Mass 1.335E8 grams (3 total)

Intact SG Volume 4,710 cubic feet Faulted SG Mass 4.45E7 grams Faulted SG Volume 1,570 cubic feet Coolant Densities Primary and Secondary water at 62.4 lbm/ft3 Table 14 - MSLB Flow Rates Flow Path Time (hour) Release Flow Note From to (lbm)

RCS to Env 0 20 - 0.134 cfm 1 RCS to Intact SGs 0 20 - 0.087 cfm RCS to Faulted SG 0 20 - 0.047 cfm Intact SGs to Env 0 2 466,400 62.3 cfm 2 2 8 1,056,000 47.0 cfm 8 20 2,112,000 47.0 cfm Faulted SG to Env 0 0.5 - 1,000 cfm 3 Flow Rate Notes:

1. RCS Leakage of 1 gpm - Volumetric leakage (gpm) from RCS is divided by 7.48 gal/ft3.

2. Intact SGs - Mass release from the intact SGs is multiplied by 1.1. Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min).

3. Faulted SG - Flow is set arbitrarily high (consistent with CLB) in order to force release of secondary side inventory within 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , consistent with a blowdown of the faulted steam generator.

CR Parameters:

Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 11.3 seconds CR Pressurization Mode Initiation Automatic at 99.3 seconds CR Ventilation System Normal Flow Rate 2,575 cfm < 11.3 seconds CR Ventilation System Filtered Makeup Rate 1,800 cfm > 99.3 seconds A7-5 to Enclosure Main Steam Line Break Analysis CR Ventilation System Recirculation Flow Rate 31,000 cfm > 99.3 seconds CR Unfiltered In-leakage 180 cfm > 11.3 seconds CR Ingress/Egress Unfiltered In-leakage 10 cfm > 0 seconds CR Ventilation System Charcoal Filter Efficiencies (Supply and Recirculation use the same filter)

All Iodine Species 99%

Particulates 99%

CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Atmospheric Dispersion Factors Time (hr) X/Q (sec/m3)

EAB* LPZ MCR Bounding Release/Receptor Location 0-2 1.80E-04 7.20E-05 7.64E-03 U1 North MSIV Room to U1 MCR Intake 2-8 - 3.30E-05 6.17E-03 U1 North MSIV Room to U1 MCR Intake 8 - 24 - 2.2E-05 2.72E-03 U2 North MSIV Room to U2 MCR Intake 24 - 96 - 9.2E-06 1.86E-03 U2 North MSIV Room to U2 MCR Intake 96 - 720 - 2.7E-06 1.52E-03 U2 North MSIV Room to U2 MCR Intake

Applied for entire duration of event in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec 24 Hours - 30 Days 2.3E-04 m3/sec A7-6 to Enclosure Steam Generator Tube Rupture Accident Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 8 Steam Generator Tube Rupture Accident Analysis A8-1 to Enclosure Steam Generator Tube Rupture Accident Analysis STEAM GENERATOR TUBE RUPTURE ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.18, Version 1 Method/Computer Program Used: RADTRAD Version 3.10 Regulatory Guidance: RG-1.183, including Appendix F Model Discussion:

The calculation was performed to address a steam generator tube rupture (SGTR). Mass transfers from the primary to secondary are calculated using the Westinghouse proprietary software tool LOFTRAN, in accordance with the VEGP current licensing basis (CLB). One steam generator (SG) tube is assumed to fail, rupturing cleanly in two. Mass transfer from the primary to the secondary continues until the break flow is terminated. Activity is released to the environment from the faulted generator until operator action is taken to isolate it. Break flow and releases are as calculated by LOFTRAN.

Primary to secondary leakage through pin-hole leaks in the SG tubes is assumed at a rate of 1 gpm (0.3 gpm to the faulted SG, 0.7 gpm to the intact SGs per the current VEGP CLB) until the SGs are isolated or no longer used for cooling. Activity is released from the other three, intact, generators through steaming via the atmospheric relief valves (ARVs) until the primary system (RCS) is reduced to cold shutdown conditions (assumed at 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ). The analysis models an exposure duration of 30 days for the LPZ and occupants of the CR.

Doses for the Pre-Accident Iodine Spike Case and the Concurrent Iodine Spike Case were calculated, with results shown below.

Results and Acceptance Limits:

Pre-Accident Concurrent Acceptance Criteria Spike Spike (Pre-Accident/Concurrent)

Location/Dose Point TEDE* (REM) TEDE* (REM) TEDE (REM)

Exclusion Area 1.6 1.4 25/2.5 Boundary (EAB)

Low Population Zone 0.9 0.8 25/2.5 (LPZ)

Main Control Room 0.6 0.5 5.0 (MCR)

Results conservatively rounded up to the nearest 0.1 Rem.

A8-2 to Enclosure Steam Generator Tube Rupture Accident Analysis Key Assumptions and Inputs:

Transient Timing Tube Rupture: Time Zero (0)

Reactor Trip: 49.7 seconds Faulted Generator Isolated 1200 seconds Break Flow Terminated 5502 seconds ARV Release from Faulted SG Ended 2162 seconds RHR In Service 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months Physical Parameters:

Parameter Value RCS Mass 2.258E08 grams RCS Volume 1.11E+04 cubic feet Intact SG Mass 1.33E+08 grams (total of 3)

Intact SG Volume 4.71E+03 cubic feet (total of 3)

Faulted SG Volume 1.57E+03 cubic feet Coolant Densities Primary and Secondary water at 62.4 lbm/ft Blowdown Sample Line Flow per SG 0.48 lbm/s Table 1 - SGTR-RCS Flow Rates RCS Transfer Pathways Description Time Flow lbm/hr Flow (CFM)

RCS Breakflow to FSG* 0-1.53 h 134067.61 35.81 RCS Leakage to FSG**,*** 0-20 h 150.16 0.04 RCS Leakage to ISG**,*** 0-20 h 350.37 0.09 Alkali Nobles (CFM) PF Iodine (CFM) PF (CFM) PF 35.81 1.00 35.81 1.00 35.81 1.00 0.04 1.00 0.04 1.00 0.04 1.00 0.09 1.00 0.09 1.00 0.09 1.00

No partitioning is modeled for breakflow to the FSG

- A conversion factor of 7.48 gal = 1ft3 was used in the conversion of the RCS leakage flows

- - RCS leakage pathway conservatively does not take credit for flashing fraction or partition factors for Iodine or Alkali Metals.

A8-3 to Enclosure Steam Generator Tube Rupture Accident Analysis Table 2 - SGTR-FSG Flow Rates FSG Transfer Pathways Description Time Flow lbm/hr Flow (CFM)

FSG to Env.**, *** 0-2 h 107000.00 28.58 FSG to Env. 2-20 h 2061.11 0.55 Alkali Nobles (CFM) PF Iodine (CFM) PF (CFM) PF 28.58 1.00 4.29 0.15 0.09 312.50 0.55 1.00 0.0055 100.00 0.0018 312.50 Description Time Flow lbm/hr Flow (CFM)

Blowdown SL to Env.* 0-20 h 1728.00 0.46 Alkali Nobles (CFM) PF Iodine (CFM) PF (CFM) PF 0.46 1.00 0.46 1.00 0.46 1.00

The blowdown sample lines discharge from the liquid region of the SG. As a result of this no partitioning of Iodine or Alkali metals is assumed to occur for this transfer pathway.

- Discharge to the environment is modeled to occur for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to remain consistent with the T/H input to dose analyses as output by LOFTRAN

- - No additional partitioning is credited for the contribution of FSG flow to the condenser. This is conservative Table 3 - SGTR-ISG Flow Rates Table 7-4: ISG to Environment Transfer Pathways Description Time Flow lbm/hr Flow (CFM)

ISG to Env. 0-2 h 348250.00 93.02 ISG to Env. 2-20 h 138644.44 37.03 Alkali Nobles (CFM) PF Iodine (CFM) PF (CFM) PF 93.02 1.00 0.93 100.00 0.30 312.50 37.03 1.00 0.37 100.00 0.12 312.50 Description Time Flow lbm/hr Flow (CFM)

Blowdown SL to Env.*,** 0-20 h 5184.00 1.38 Alkali Nobles (CFM) PF Iodine (CFM) PF (CFM) PF 1.38 1.00 1.38 1.00 1.38 1.00

The blowdown sample lines discharge from the liquid region of the SG. As a result of this no partitioning of Iodine or Alkali metals is assumed to occur for this transfer pathway.

- The blowdown sample line flows for the ISG is 3 times that of the FSG.

A8-4 to Enclosure Steam Generator Tube Rupture Accident Analysis Radioactivity Considerations:

No fuel failure occurs as a result of the SGTR.

Iodine Release Species: 97% elemental, 3% organic.

Initial RCS activity I-131 through I-135 Pre-Accident Iodine Spike Case 60 µCi/gm DEI-131 Concurrent Accident Spike 1.0 µCi/gm DEI-131 Noble Gases other than Xe-133m 280 µCi/gm DE Xe-133 Other Noble Gases, Halogens, and Alkali Metals 1% clad defects Concurrent Accident Spike Appearance 335 times the Normal Rate for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Table 4 - Normal RCS Iodine Concentrations RCS Source Term RCS Source Term Nuclide

(µCi/gm) (Ci)

I-130 2.19E-02 4.938E+00 I-131 7.61E-01 1.717E+02 I-132 7.79E-01 1.760E+02 I-133 1.16E+00 2.615E+02 I-134 1.66E-01 3.744E+01 I-135 6.40E-01 1.446E+02 Table 4 Notes:

1. Normal RCS Iodine source term concentrations multiplied by RCS mass of 2.258E+08 grams and 1E-06Ci/µCi to obtain the specific activity.

Table 5 - Normal Iodine Appearance Rates Radionuclide Appearance Rate (Ci/hr)

I-131 1.75E+01 I-132 7.05E+01 I-133 3.44E+01 I-134 3.33E+01 I-135 2.95E+01 Table 5 Notes:

1. The Concurrent spike case multiplies the normal values in Table 5 by 335 and injects into the RCS over an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

A8-5 to Enclosure Steam Generator Tube Rupture Accident Analysis Initial Iodine Activities in RCS and Secondary Coolant The initial iodine activities in the RCS and the secondary coolant corresponding to 60, 1.0, and 0.1 Ci/g DEI are shown in the following table. The normal RCS activity is used as the starting point for the concurrent spike case.

Table 6 - RCS Source Terms for Pre-Accident and Concurrent Iodine Spike Cases Nuclide RCS RCS Pre- FSG (Ci) ISG (Ci)

(Normal) Accident (Ci) Spike (Ci)

Kr-83m 1.06E+02 1.06E+02 0.00E+00 0.00E+00 Kr-85 8.68E+02 8.68E+02 0.00E+00 0.00E+00 Kr-85m 1.69E+02 1.69E+02 0.00E+00 0.00E+00 Kr-87 1.10E+02 1.10E+02 0.00E+00 0.00E+00 Kr-88 3.70E+02 3.70E+02 0.00E+00 0.00E+00 Xe-131m 3.30E+02 3.30E+02 0.00E+00 0.00E+00 Xe-133 2.60E+04 2.60E+04 0.00E+00 0.00E+00 Xe-133m 3.60E+02 3.60E+02 0.00E+00 0.00E+00 Xe-135 7.85E+02 7.85E+02 0.00E+00 0.00E+00 Xe-135m 4.94E+01 4.94E+01 0.00E+00 0.00E+00 Xe-138 6.25E+01 6.25E+01 0.00E+00 0.00E+00 Br-82 1.16E+00 1.16E+00 2.28E-02 6.83E-02 Br-83 2.23E+01 2.23E+01 4.39E-01 1.32E+00 Br-84 1.05E+01 1.05E+01 2.07E-01 6.21E-01 I-130 4.94E+00 4.94E+00 9.72E-02 2.92E-01 I-131 1.72E+02 1.03E+04 3.38E+00 1.01E+01 I-132 1.76E+02 1.06E+04 3.46E+00 1.04E+01 I-133 2.62E+02 1.57E+04 5.15E+00 1.54E+01 I-134 3.75E+01 2.25E+03 7.37E-01 2.21E+00 I-135 1.45E+02 8.68E+03 2.85E+00 8.54E+00 Cs-134 4.49E+02 4.49E+02 8.84E+00 2.65E+01 Cs-134m 5.55E+00 5.55E+00 1.09E-01 3.28E-01 Cs-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs-136 6.14E+02 6.14E+02 1.21E+01 3.62E+01 Cs-137 4.34E+02 4.34E+02 8.54E+00 2.56E+01 Cs-138 2.41E+02 2.41E+02 4.74E+00 1.42E+01 Rb-86 6.05E+00 6.05E+00 1.19E-01 3.57E-01 Rb-88 9.77E+02 9.77E+02 1.92E+01 5.77E+01 Rb-89 2.28E+01 2.28E+01 4.49E-01 1.35E+00 A8-6 to Enclosure Steam Generator Tube Rupture Accident Analysis Table 7 - Feedwater Source Terms Nuclide FSG ISG (Ci/hr) (Ci/hr)

I-130 1.05E+00 1.65E-01 I-131 3.64E+01 5.72E+00 I-132 3.73E+01 5.86E+00 I-133 5.54E+01 8.71E+00 I-134 7.93E+00 1.25E+00 I-135 3.06E+01 4.82E+00 Table 7 Notes:

1. Activity injection from main feedwater is modeled to stop at the time of Rx trip (49.7 seconds).

CR Parameters:

Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 121 seconds CR Pressurization Mode Initiation Automatic at 211 seconds CR Ventilation System Normal Flow Rate 2,575 cfm < 121 seconds CR Ventilation System Filtered Makeup Rate 1,800 cfm > 121 seconds CR Ventilation System Recirculation Flow Rate 31,000 cfm > 121seconds CR Unfiltered In-leakage 180 cfm > 121 seconds CR Ingress/Egress Unfiltered In-leakage 10 cfm > 0 seconds CR Ventilation System Charcoal Filter Efficiencies (Supply and Recirculation use the same filter)

All Iodine Species 99%

Particulates 99%

CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 A8-7 to Enclosure Steam Generator Tube Rupture Accident Analysis Atmospheric Dispersion Factors Time (hr) X/Q (sec/m3)

EAB* LPZ MCR Bounding Release/Receptor Location 0-2 1.80E-04 7.20E-05 7.64E-03 U1 North MSIV Room to U1 MCR Intake 2-8 - 3.30E-05 6.17E-03 U1 North MSIV Room to U1 MCR Intake 8 - 24 - 2.2E-05 2.72E-03 U2 North MSIV Room to U2 MCR Intake 24 - 96 - 9.2E-06 1.86E-03 U2 North MSIV Room to U2 MCR Intake 96 - 720 - 2.7E-06 1.52E-03 U2 North MSIV Room to U2 MCR Intake

Applied for entire duration of event in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec 24 Hours - 30 Days 2.3E-04 m3/sec A8-8 to Enclosure Control Rod Ejection Accident Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 9 Control Rod Ejection Accident Analysis A9-1 to Enclosure Control Rod Ejection Accident Analysis CONTROL ROD EJECTION ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.20, Version 2 Method/Computer Program Used: RADTRAD, version 3.10 Regulatory Guidance: RG-1.183 Rev. 0, including Appendix H Model Discussion The calculation was performed to address a Control Rod Ejection Accident (CREA). The scenario for the CREA is that the reactivity excursion due to a control rod ejection leads to localized fuel damage (including a small fraction of fuel melt). The local fuel damage results in increased radioactivity in the Reactor Coolant System (RCS). Activity in the steam generators (SG) due to primary-to-secondary leakage is released to the environment via steaming until the RCS is placed on RHR cooling.

Two release pathways are considered independently, in accordance with RG-1.183:

Containment Leakage - Activity from fuel melting, fuel cladding damage, and initial RCS inventory instantaneously reaches the containment at the onset of the accident and is available for release to the environment.

Secondary System Release - Activity from fuel melting, fuel cladding damage, and initial RCS inventory instantaneously mixes in the RCS at the onset of the accident and is available for release to the secondary system and eventually to the environment.

Results and Acceptance Limits Location Dose (Rem TEDE)

Containment Secondary Limit Release Side Release EAB 1.5 0.5 6.3 LPZ 1.8 0.4 6.3 Control Room 0.6 1.1 5 (Note that rounding is applied to all values)

A9-2 to Enclosure Control Rod Ejection Accident Analysis Key Assumptions and Inputs Physical Parameters Parameter Value Reactor Power Level 3636 MWt (includes uncertainty)

Containment Volume 2.93E+6 ft3 Containment Leakage 0.21% per day for first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.105% per day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Particulate Removal 3.005E-2 per hour, credit is taken for natural deposition in Containment RCS Mass 2.258E+8 grams RCS Volume 1.11E+4 ft3 SG Mass 391,908 lbm (all 4 SGs)

SG Volume 6280 ft3 (all 4 SGs)

Coolant Densities Primary and Secondary water at 62.4 lbm/ft3 Partition Factors Iodine PF = 100, Alkali Metals PF = 312 (moisture carryover

= 0.32%)

Noble Gases PF = 1 Primary to Secondary Leakage 1 gpm total, for the first 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months of the accident.

Secondary System Mass releases 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.5E+5 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.365E+6 lbm 8 - 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months 2.73E+6 lbm A9-3 to Enclosure Control Rod Ejection Accident Analysis Table 1 - Secondary Side Pathway Flow Rates Flow Path Time (hour) Flow Note From to RCS to SGs 0 20 0.134 cfm 1 RCS to Env (Noble Gases) 0 20 0.134 cfm 1, 2 0 2 80.8 cfm 3 SGs to Env 2 8 66.8 cfm 8 20 66.8 cfm Flow Rate Notes:

1. RCS Leakage of 1 gpm - Volumetric leakage (gpm) from RCS is divided by 7.48 gal/ft3.

2. The noble gases are released without holdup or retention, modeled as a direct release from the RCS to the environment at the primary-to-secondary leakage rate.

3. SGs - Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min).

Radioactivity Considerations 0.25% of Fuel Rods experience melting.

100% of the noble gases and 25% of the Iodine isotopes (containment) or 50% of the iodine isotopes (secondary side release) within the melting rods are available for release.

10% of the fuel rods experience cladding failure. A radial power peaking factor of 1.7 is applied to the damaged rods.

The fractions of fission product inventory contained within the fuel rod gaps are:

o Iodine isotopes and Noble gases 0.10 o Other Halogens 0.05 o Alkali Metals 0.12 Core Fission product inventories are taken from an equilibrium cycle based upon a power level of 3636 MWt. To account for potential cycle-to-cycle variations, a 10%

fuel design margin is applied.

Containment Release - 100% of the activity released from the core due to fuel melting and cladding failure and the initial RCS activity is instantaneously released to and uniformly mixed in the containment at the onset of the accident.

Secondary Side Release - 100% of the activity released from the core due to fuel melting and cladding failure and the initial RCS and secondary side activities are instantaneously mixed within the RCS at the onset of the accident.

Chemical form of iodine released to containment is 95% particulate, 4.85 elemental, and 0.15% organic.

Chemical form of iodine released through the SGs is 97% elemental and 3% organic.

RCS non-iodine activity (alkali metals and other halogens) includes an assumption of normal operations 1% fuel clad defects.

The radioiodine concentration in the RCS is assumed to be at the Technical A9-4 to Enclosure Control Rod Ejection Accident Analysis Specification equilibrium limit of 1.0 µCi/gm DE I-131.

The noble gas concentration in the RCS is based on 280 µCi/gm DE Xe-133.

The radioiodine concentration in the secondary system is assumed to be at the Technical Specification limit of 0.1 µCi/gm DE I-131.

The concentrations of Alkali Metals in Secondary are based upon a ratio of the concentration in the RCS: Given 0.1 µCi/gm DE I-131 in the secondary and 1.0 µCi/gm in the RCS, the concentrations of alkali metals in the secondary are assumed to be 10% of those in the RCS.

A9-5

Attachment 9 to Enclosure Control Rod Ejection Accident Analysis Containment and RCS Activities Table 2 reports the total activity released to containment for the containment release case and Table 3 reports the total activity in the RCS for the secondary side release case. Table 4 reports the activity initially in the SGs prior to the event, which is available for release in the secondary side case.

Table 2 - Containment Release Activities Nuclide Core Radial Gap Fraction Melt Fraction Fuel Release RCS Source Total Release Source Peaking Release of Fuel Release of Fuel Source Term Term (Ci) to Containment Term Factor Fraction with Clad Fraction Melted (Ci) (Ci)

(Ci) Damage Kr-83m 1.29E+07 0.1 1.0 2.740E+05 1.059E+02 2.741E+05 Kr-85 1.12E+06 0.1 1.0 2.384E+04 8.669E+02 2.471E+04 Kr-85m 2.74E+07 0.1 1.0 5.823E+05 1.684E+02 5.824E+05 Kr-87 5.40E+07 0.1 1.0 1.148E+06 1.100E+02 1.148E+06 Kr-88 7.23E+07 0.1 1.0 1.536E+06 3.696E+02 1.536E+06 Xe-131m 1.42E+06 0.1 1.0 3.011E+04 3.292E+02 3.044E+04 Xe-133 2.15E+08 0.1 1.0 4.574E+06 2.594E+04 4.600E+06 Xe-133m 6.83E+06 0.1 1.0 1.452E+05 3.597E+02 1.455E+05 Xe-135 5.10E+07 0.1 1.0 1.084E+06 7.838E+02 1.085E+06 Xe-135m 4.59E+07 0.1 1.0 9.757E+05 4.940E+01 9.757E+05 Xe-138 1.87E+08 0.1 1.0 3.976E+06 6.245E+01 3.976E+06 Br-82 3.41E+05 0.05 0.0 2.898E+03 1.157E+00 2.899E+03 Br-83 1.28E+07 0.05 0.0 1.086E+05 2.230E+01 1.087E+05 Br-84 2.32E+07 0.05 0.0 1.974E+05 1.051E+01 1.974E+05 I-130 2.04E+06 1.70 0.1 0.1 0.25 0.0025 3.676E+04 4.938E+00 3.676E+04 I-131 1.07E+08 0.1 0.25 1.938E+06 1.718E+02 1.938E+06 I-132 1.57E+08 0.1 0.25 2.829E+06 1.759E+02 2.829E+06 I-133 2.20E+08 0.1 0.25 3.972E+06 2.619E+02 3.972E+06 I-134 2.47E+08 0.1 0.25 4.459E+06 3.748E+01 4.459E+06 I-135 2.10E+08 0.1 0.25 3.787E+06 1.445E+02 3.787E+06 Cs-134 1.67E+07 0.12 0.0 3.415E+05 4.491E+02 3.420E+05 Cs-134m 4.50E+06 0.12 0.0 9.176E+04 5.549E+00 9.176E+04 Cs-135 4.51E+01 0.12 0.0 9.200E-01 0.000E+00 9.200E-01 Cs-136 5.37E+06 0.12 0.0 1.096E+05 6.135E+02 1.102E+05 Cs-137 1.18E+07 0.12 0.0 2.399E+05 4.337E+02 2.403E+05 Cs-138 2.04E+08 0.12 0.0 4.167E+06 2.409E+02 4.167E+06 Rb-86 2.07E+05 0.12 0.0 4.230E+03 6.041E+00 4.236E+03 Rb-88 7.36E+07 0.12 0.0 1.501E+06 9.766E+02 1.502E+06 Rb-89 9.64E+07 0.12 0.0 1.966E+06 2.282E+01 1.966E+06 A9-6

Attachment 9 to Enclosure Control Rod Ejection Accident Analysis Table 3 - RCS Activities for Secondary Side Release Nuclide Core Radial Gap Fraction Melt Fraction Fuel Release RCS Source Total Source Peaking Release of Fuel Release of Fuel Source Term Term (Ci) Release to Term Factor Fraction with Clad Fraction Melted (Ci) RCS (Ci)

(Ci) Damage Kr-83m 1.29E+07 0.1 1.0 2.740E+05 1.059E+02 2.741E+05 Kr-85 1.12E+06 0.1 1.0 2.384E+04 8.669E+02 2.471E+04 Kr-85m 2.74E+07 0.1 1.0 5.823E+05 1.684E+02 5.824E+05 Kr-87 5.40E+07 0.1 1.0 1.148E+06 1.100E+02 1.148E+06 Kr-88 7.23E+07 0.1 1.0 1.536E+06 3.696E+02 1.536E+06 Xe-131m 1.42E+06 0.1 1.0 3.011E+04 3.292E+02 3.044E+04 Xe-133 2.15E+08 0.1 1.0 4.574E+06 2.594E+04 4.600E+06 Xe-133m 6.83E+06 0.1 1.0 1.452E+05 3.597E+02 1.455E+05 Xe-135 5.10E+07 0.1 1.0 1.084E+06 7.838E+02 1.085E+06 Xe-135m 4.59E+07 0.1 1.0 9.757E+05 4.940E+01 9.757E+05 Xe-138 1.87E+08 0.1 1.0 3.976E+06 6.245E+01 3.976E+06 Br-82 3.41E+05 0.05 0.0 2.898E+03 1.157E+00 2.899E+03 Br-83 1.28E+07 0.05 0.0 1.086E+05 2.230E+01 1.087E+05 Br-84 2.32E+07 0.05 0.0 1.974E+05 1.051E+01 1.974E+05 I-130 2.04E+06 1.70 0.1 0.1 0.5 0.0025 3.892E+04 4.938E+00 3.892E+04 I-131 1.07E+08 0.1 0.5 2.052E+06 1.718E+02 2.052E+06 I-132 1.57E+08 0.1 0.5 2.996E+06 1.759E+02 2.996E+06 I-133 2.20E+08 0.1 0.5 4.205E+06 2.619E+02 4.206E+06 I-134 2.47E+08 0.1 0.5 4.721E+06 3.748E+01 4.721E+06 I-135 2.10E+08 0.1 0.5 4.010E+06 1.445E+02 4.010E+06 Cs-134 1.67E+07 0.12 0.0 3.415E+05 4.491E+02 3.420E+05 Cs-134m 4.50E+06 0.12 0.0 9.176E+04 5.549E+00 9.176E+04 Cs-135 4.51E+01 0.12 0.0 9.200E-01 0.000E+00 9.200E-01 Cs-136 5.37E+06 0.12 0.0 1.096E+05 6.135E+02 1.102E+05 Cs-137 1.18E+07 0.12 0.0 2.399E+05 4.337E+02 2.403E+05 Cs-138 2.04E+08 0.12 0.0 4.167E+06 2.409E+02 4.167E+06 Rb-86 2.07E+05 0.12 0.0 4.230E+03 6.041E+00 4.236E+03 Rb-88 7.36E+07 0.12 0.0 1.501E+06 9.766E+02 1.502E+06 Rb-89 9.64E+07 0.12 0.0 1.966E+06 2.282E+01 1.966E+06 Tables 2 and 3 notes:

1. Fuel Release Source Term = Core Source Term x Radial Peaking Factor x [(Gap Release Fraction x Fraction of Fuel with Clad Damage) + (Melt Release Fraction x Fraction of Fuel Melted)]

2. RCS source term is based on 1.0 µCi/gm DE I-131 (I-131 to I-135), 280 µCi/gm DE Xe-133 for noble gases (excluding Kr-83m), and 1% defective fuel for other nuclides.

3. Total Activity - This is the sum of the fuel release source term and RCS source term.

A9-7 to Enclosure Control Rod Ejection Accident Analysis Table 4 - Initial SG Activities Nuclide Secondary Side Source Term (Ci)

Br-82 9.112E-02 Br-83 1.756E+00 Br-84 8.277E-01 I-130 3.888E-01 I-131 1.352E+01 I-132 1.386E+01 I-133 2.059E+01 I-134 2.948E+00 I-135 1.139E+01 Cs-134 3.536E+01 Cs-134m 4.369E-01 Cs-135 0.000E+00 Cs-136 4.831E+01 Cs-137 3.415E+01 Cs-138 1.897E+01 Rb-86 4.757E-01 Rb-88 7.690E+01 Rb-89 1.797E+00 Table 4 note:

1. Secondary side source term is based on 10% of the equilibrium RCS source term.

A9-8 to Enclosure Control Rod Ejection Accident Analysis Control Room Ventilation Parameters Parameter Value CR Volume 149,000 ft3 CR Isolation Automatic at 131 Seconds CR Pressurization Mode Initiation Automatic at 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 Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 A9-9 to Enclosure Control Rod Ejection Accident Analysis Atmospheric Dispersion Factors Table 5 - Atmospheric Dispersion Factors X/Q (sec/m3)

EAB* LPZ Containment Secondary Time (hr) Bounding Release &

Release to Side Release Receptor Location MCR to MCR 0-2 1.80E-04 7.20E-05 2.22E-03 7.64E-03 U1 N MSIV to U1 MCR 2-8 - 3.30E-05 1.55E-03 6.17E-03 U1 N MSIV to U1 MCR 8 - 24 - 2.2E-05 6.57E-04 2.72E-03 U2 N MSIV to U2 MCR 24 - 96 - 9.2E-06 5.80E-04 1.86E-03 U2 N MSIV to U2 MCR 96 - 720 - 2.7E-06 4.47E-04 1.52E-03 U2 N MSIV to U2 MCR

Applied for entire duration of event in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec 24 Hours - 30 Days 2.3E-04 m3/sec A9-10 0 to Enclosure Locked Rotor Accident Analysis Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 10 Locked Rotor Accident Analysis A10-1 0 to Enclosure Locked Rotor Accident Analysis LOCKED ROTOR ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: X6CAJ.19, Version 2 Method/Computer Program Used: RADTRAD, Version 3.10 Regulatory Guidance: RG-1.183, including Appendix G Model Discussion The calculation was performed to address a Locked Rotor Accident (LRA). The scenario for the LRA is that a reactor coolant pump rotor is postulated to seize, leading to reduced coolant flow and reactor trip. The transient causes fuel damage, resulting in increased radioactivity in the Reactor Coolant System (RCS). Activity in the steam generators (SG) due to primary-to-secondary leakage is released to the environment via steaming until the RCS is placed on RHR cooling.

Results and Acceptance Limits:

Location Dose (Rem TEDE)

Calculated Limit EAB <0.1 2.5 LPZ <0.1 2.5 Control Room 0.3 5 Key Assumptions and Inputs Physical Parameters Parameter Value Reactor Power Level: 3636 MWt (includes uncertainty)

RCS Mass: 2.258E8 grams RCS Volume: 1.11E4 cubic feet SG Mass: 1.78E8 lbm (all 4 SGs)

SG Volume: 6,281 cubic feet (all 4 SGs)

Secondary System Margin 10% increase is added to mass flows Coolant Densities: Primary and Secondary water at 62.4 lbm/ft3 Partition Factors: Iodine PF = 100 Alkali Metals PF = 312 (moisture carryover

= 0.32%)

Primary to Secondary Leakage: 1 gpm total.

A10-2 0 to Enclosure Locked Rotor Accident Analysis Table 1: Flow Rates After 10% Margin Adjustment Pathway Time Release (lbm) Flow Note From To RCS to SG 0 8 - 0.134 cfm 1 SG to 0 2 610,500 81.5 cfm 2 Environment 2 8 1,501,500 66.8 cfm 8 20 3,003,000 66.8 cfm Flow Rate Notes:

1. RCS - Volumetric leakage (1 gallons/minute) from the RCS is divided by 7.48 gal/ft3.

1. SG - The mass release from the SG is increased by 10% for margin. The flow is then the mass release (lbm) divided by 62.4 lbm/ft3 and divided by the time duration (min).

Radioactivity Considerations:

5% of the fuel rods experience cladding failure. A radial power peaking factor of 1.7 is applied to the damaged rods.

The fractions of fission product inventory contained within the fuel rod gaps are:

o I-131 0.08 o Kr-85 0.10 o Other Halogens and Noble Gases 0.05 o Alkali Metals 0.12 Core Fission product inventories are taken from an equilibrium cycle based upon a power level of 3636 MWt. To account for potential cycle-to-cycle variations, 10%

margins are applied to the core inventory.

100% of the activity released from the core due to cladding failure is instantaneously mixed within the RCS at the onset of the accident.

Chemical form of iodine released to the environment from the steam generators is 97%

elemental, and 3% organic. The removal mechanism for this pathway is the same for all chemical forms of iodine.

Radial peaking factor for rods with cladding damage is assumed to be 1.7.

The initial RCS activity represents normal operations with a 1.0 µCi/g DEI-131 Technical Specification limit, a 280 µCi/g DEXe-133 Technical Specification limit, and 1% clad defects for all other radionuclides.

The initial radioiodine concentration in the secondary system is assumed to be at the Technical Specification limit of 0.1 µCi/gm DEI The initial concentrations of Alkali Metals in Secondary are based upon a ration of the concentration in the RCS: Given 0.1 µCi/g DEI in the secondary and 1.0 µCi/g in the RCS, the concentrations of alkali metals in the secondary are assumed to be 10% of those in the RCS.

Noble gases are not retained on the secondary side.

A10-3 0 to Enclosure Locked Rotor Accident Analysis RCS Activities Table 2 - RCS Activities RG Total RCS Nuclide 1.183 Source Term Group (Ci)

Kr-83m NG 5.490E+04 Kr-85 NG 1.040E+04 Kr-85m NG 1.166E+05 Kr-87 NG 2.297E+05 Kr-88 NG 3.076E+05 Xe-131m NG 6.351E+03 Xe-133 NG 9.408E+05 Xe-133m NG 2.939E+04 Xe-135 NG 2.176E+05 Xe-135m NG 1.952E+05 Xe-138 NG 7.953E+05 Br-82 H 1.450E+03 Br-83 H 5.435E+04 Br-84 H 9.870E+04 I-130 H 8.654E+03 I-131 H 7.297E+05 I-132 H 6.659E+05 I-133 H 9.348E+05 I-134 H 1.049E+06 I-135 H 8.912E+05 Cs-134 AM 1.712E+05 Cs-134m AM 4.588E+04 Cs-135 AM 4.600E-01 Cs-136 AM 5.540E+04 Cs-137 AM 1.204E+05 Cs-138 AM 2.084E+06 Rb-86 AM 2.121E+03 Rb-88 AM 7.514E+05 Rb-89 AM 9.832E+05 Sb-124 TG 0.000E+00 Sb-125 TG 0.000E+00 Sb-126 TG 0.000E+00 Sb-127 TG 0.000E+00 Sb-129 TG 0.000E+00 Te-125m TG 0.000E+00 A10-4 0 to Enclosure Locked Rotor Accident Analysis RG Total RCS Nuclide 1.183 Source Term Group (Ci)

Te-127 TG 0.000E+00 Te-127m TG 0.000E+00 Te-129 TG 0.000E+00 Te-129m TG 0.000E+00 Te-131 TG 0.000E+00 Te-131m TG 0.000E+00 Te-132 TG 0.000E+00 Te-133 TG 0.000E+00 Te-133m TG 0.000E+00 Te-134 TG 0.000E+00 Sr-89 TG 0.000E+00 Sr-90 TG 0.000E+00 Sr-91 TG 0.000E+00 Sr-92 TG 0.000E+00 Ba-137m TG 0.000E+00 Ba-139 TG 0.000E+00 Ba-140 TG 0.000E+00 Ba-141 TG 0.000E+00 Mo-99 NM 0.000E+00 Pd-109 NM 0.000E+00 Rh-103m NM 0.000E+00 Rh-105 NM 0.000E+00 Rh-106 NM 0.000E+00 Ru-103 NM 0.000E+00 Ru-105 NM 0.000E+00 Ru-106 NM 0.000E+00 Tc-99 NM 0.000E+00 Tc-99m NM 0.000E+00 Ce-141 CE 0.000E+00 Ce-143 CE 0.000E+00 Ce-144 CE 0.000E+00 Np-237 CE 0.000E+00 Np-238 CE 0.000E+00 Np-239 CE 0.000E+00 Pu-238 CE 0.000E+00 Pu-239 CE 0.000E+00 Pu-240 CE 0.000E+00 Pu-241 CE 0.000E+00 Pu-242 CE 0.000E+00 Pu-243 CE 0.000E+00 A10-5 0 to Enclosure Locked Rotor Accident Analysis RG Total RCS Nuclide 1.183 Source Term Group (Ci)

Am-241 LA 0.000E+00 Am-242 LA 0.000E+00 Am-243 LA 0.000E+00 Cm-242 LA 0.000E+00 Cm-244 LA 0.000E+00 Eu-154 LA 0.000E+00 Eu-155 LA 0.000E+00 Eu-156 LA 0.000E+00 La-140 LA 0.000E+00 La-141 LA 0.000E+00 La-142 LA 0.000E+00 La-143 LA 0.000E+00 Nb-95 LA 0.000E+00 Nb-95m LA 0.000E+00 Nb-97 LA 0.000E+00 Nb-97m LA 0.000E+00 Nd-147 LA 0.000E+00 Pm-147 LA 0.000E+00 Pm-148 LA 0.000E+00 Pm-148m LA 0.000E+00 Pm-149 LA 0.000E+00 Pm-151 LA 0.000E+00 Pr-143 LA 0.000E+00 Pr-144 LA 0.000E+00 Pr-144m LA 0.000E+00 Sm-151 LA 0.000E+00 Sm-153 LA 0.000E+00 Y-90 LA 0.000E+00 Y-91 LA 0.000E+00 Y-91m LA 0.000E+00 Y-92 LA 0.000E+00 Y-93 LA 0.000E+00 Y-95 LA 0.000E+00 Zr-95 LA 0.000E+00 Zr-97 LA 0.000E+00 Containment and RCS Activities Notes:

1. The Total RCS source term is the core gap release activity consistent with 5% fuel damage plus the RCS initial activity.

A10-6 0 to Enclosure Locked Rotor Accident Analysis Initial Iodine Activities in the Secondary System The initial iodine activities in the secondary coolant corresponding to 0.1 µCi/g DEI are shown in the following table.

Table 3 - Initial Iodine Activities in the Secondary System Isotope Secondary System Initial Activity (µCi/g)

I-131 7.61E-02 I-132 7.79E-02 I-133 1.16E-01 I-134 1.66E-02 I-135 6.40E-02 Alkali Metals in the Secondary System The initial concentrations are assumed to be 10% of the RCS initial activities.

Table 4 - Alkali Metals in the Secondary System Isotope Concentration (µCi/g) Activity (Curies)

Cs-134 1.99E-01 3.540E+01 Cs-134m 2.46E-03 4.374E-01 Cs-135 0.00E+00 0.000E+00 Cs-136 2.72E-01 4.836E+01 Cs-137 1.92E-01 3.419E+01 Cs-138 1.07E-01 1.899E+01 Rb-86 2.68E-03 4.762E-01 Rb-88 4.33E-01 7.699E+01 Rb-89 1.01E-02 1.799E+00 Notes: 1 Initial Alkali Metals in the Secondary System Notes:

1. Secondary Activities - the concentrations in Column 2 are multiplied by the mass of 1.78E8 grams and by 1.0E-06 Ci/µCi to obtain the activity in curies in Column 3.

Control Room Ventilation Parameters CR Volume 149,000 ft3 CR Isolation Automatic at 608 Seconds CR Pressurization Mode Initiation Automatic at 698 Seconds CR Ventilation System Normal Flow Rate 2,575 cfm < 608 seconds CR Ventilation System Filtered Makeup Rate 1,800 cfm > 698 seconds CR Ventilation System Recirculation Flow Rate 31,000 cfm > 698 seconds CR Ventilation System Charcoal Filter Efficiencies (Supply and Recirculation use the same filter)

A10-7 0 to Enclosure Locked Rotor Accident Analysis 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 Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Atmospheric Dispersion Factors Table 6 - Atmospheric Dispersion Factors Time (hr) EAB* LPZ MCR Bounding Release & Receptor Location 0-2 1.80E-4 7.20E-5 7.64E-3 U1 N MSIV to U1 MCR 2-8 3.30E-5 6.17E-3 U1 N MSIV to U1 MCR 8 - 24 2.20E-5 2.72E-3 U2 N MSIV to U2 MCR 24 - 96 9.20E-6 1.86E-3 U2 N MSIV to U2 MCR 96 - 720 2.70E-6 1.52E-3 U2 N MSIV to U2 MCR

Applied for entire duration of event in analysis EAB & LPZ Breathing Rates 0-8 Hours 3.5E-04 m3/sec 8-24 Hours 1.8E-04 m3/sec 24 Hours - 30 Days 2.3E-04 m3/sec A10-8 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Vogtle Electric Generating Plant - Units 1 & 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 11 VEGP AST Accident Analysis Input Values Comparison Tables A11-1 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables VEGP AST Accident Analysis Input Values Comparison Tables To facilitate the review and to more readily assess the impact of the adoption of the Alternative Source Term (AST) at Vogtle Nuclear Plant (VEGP), summary tables are provided in this enclosure for each accident being analyzed including a comparison between current licensing basis (CLB) input parameters and the values utilized in the new AST accident analysis, and the basis for any changes. The tables are provided within this enclosure for the following accident scenarios:

Table 2 - Loss of Coolant Accident (LOCA)

Table 3 - Fuel Handling Accident (FHA)

Table 4 - Main Steam Line Break (MSLB) Accident Table 5 - Steam Generator Tube Rupture (SGTR) Accident Table 6 - Control Rod Ejection Accident Table 7 - Locked Rotor Accident Additionally, Table 1, "Control Room Parameters," is provided to show the parameters of interest for Control Room habitability. In this table, the LOCA parameters are provided as they resulted in the most limiting dose to the Control Room occupants.

Table 1: Control Room Parameters Input/Assumption CLB Value New AST Value Reason for Change Control Room Volume 1.72E+05 ft3 1.49E+05 ft3 AST reduces CR free volume to account for equipment.

Normal Operation Filtered Make-up Flow 0 cfm 0 cfm No change Rate Filtered Recirculation 0 cfm 0 cfm No change Flow Rate Unfiltered Make-up Flow 3000 cfm 2575 cfm AST value based on process flow diagrams biased Rate conservatively to account for uncertainty (TS 5.5.11).

Unfiltered In-leakage 0 cfm 0 cfm No change Emergency Operation Recirculation Mode Filtered Make-up Flow 1500 cfm 1800 cfm AST value based on CREFS process flow diagrams Rate biased conservatively to account for uncertainty (TS 5.5.11).

Filtered Recirculation 19000 cfm 31000 cfm AST considers dual unit actuation resulting from Flow Rate high radiation signal.

Unfiltered Make-up Flow 0 cfm 0 cfm No change Rate Unfiltered In-leakage 140 cfm 190 cfm The revised value is intended to provide operational margin to the CR measured CR in-leakage. Includes 10 cfm for CR ingress/egress.

Filter Efficiencies Pressurization Filters All iodine99% All iodine 99% No change Recirculation Filters Elemental - 99% Elemental - 99% No change Organic - 99% Organic - 99%

Particulate - 99% Particulate - 99%

Particulate 99% 99% No change Occupancy 0-24 hours 100% 100% No change 1-4 days 60% 60%

4-30 days 40% 40%

A11-2 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Table 1: Control Room Parameters Input/Assumption CLB Value New AST Value Reason for Change Breathing Rate 3.47E-4 m3/sec (0-720 hr) 3.5E-4 m3/sec (0-720 hr) Difference complies with RG-1.183 Rev. 0 Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Containment Purge Iodine Chemical Form 5% particulate, 91% elemental, 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

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% elemental, 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

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

Containment Spray Stop 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months No change Time Containment Spray 2500 gpm 2500 gpm No change Flow Rate Elemental Iodine Spray 10 hr-1 13.7 hr-1 Revision is consistent with RG 1.183 Appendix A Removal Coefficient RP 3.3 Aerosol Spray Removal 4.2 hr-1 5.34 hr-1 Revision is consistent with RG 1.183 Appendix A Coefficient 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.2%/day 0.21%/day 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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% elemental, The revised percentages are as specified in RG 4% organic 3% organic 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 Start 30 minutes 30 minutes No change Time ECCS Leakage Flow Rate 2 gpm 2 gpm No change ECCS Flashing Fraction 10% 10% No change ECCS Leakage to the RWST (Not Explicitly Modeled in the CLB)

A11-3 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Table 3: FHA Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Iodine Chemical Form 0% aerosol, 99.75% elemental, 0% aerosol, 99.85% Chemical composition is as described in RG 1.183 0.25% organic elemental, 0.15% organic Appendix B Section 2.

Number of Fuel 314 (1 FA + 50 rods in target 314 Rods (1 FA + 50 rods in No change Assemblies Damaged assembly) target assembly)

Percentage of Fuel 100% 100% No change Rods Damaged No. of rods exceeding 0 314 100% of rods analyzed with elevated gap 6.3 kw/ft above 54 fractions for conservatism. Requesting exception GWD/MTU for 40% of rods.

Water Level Above 23 ft 23 ft No change Damaged Fuel Pool Decontamination 200-Overall 200-Overall Decontamination Factors are as described in RG Factors 1.183 Appendix B Section 2.

Delay Before Fuel 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months AST analyzed at earlier times Movement Onsite X/Qs AST did not analyze containment as the release Containment durations being equal the fuel building will be 0-2 hrs 1.04E-3 sec/m3 N/A bounding as the X/Qs are nearly 6 times larger 2-8 hrs 1.04E-3 sec/m3 N/A throughout the release duration. Regarding the 8-24 hrs 1.04E-3 sec/m3 N/A Fuel Building, the AST X/Qs are bounding Fuel Building throughout the entire release duration which is 0-2 hrs 4.99E-3 sec/m3 6.01E-3 sec/m3 conservative.

2-8 hrs 4.99E-3 sec/m3 4.44E-3 sec/m3 8-24 hrs 4.99E-3 sec/m3 1.71E-3 sec/m3 Table 4: MSLB Accident Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Maximum Pre-Accident Maximum Value in TS 60 Ci/gm Dose Equivalent Specified by RG-1.183 Section E.

Iodine Spike I-131 Concentration Concurrent Iodine 500 X Equilibrium 500 X Equilibrium No change Spike Appearance Rate Initial Steam Generator 0.1 Ci/gf Dose Equivalent 0.1 Ci/gf Dose Equivalent No change Iodine Source Term Iodine Iodine Iodine Chemical Form 100% Elemental 0% aerosol, 97% elemental, The AST chemical form is as provided in RG 1.183 3% organic Appendix E, Section 4.

Percentage of Fuel 0% 0% No change. Note- the MSLB does not result in Rods Failed failed fuel. This is a leaking fuel pre-condition included for conservatism.

RCS Mass 2.53E+08 g 2.258E+08 g RCS mass as provided by NSSS vendor Steam Generator 1.9E+08 (all 4 SGs) 1.78E+08 g (all 4 SGs) SG mass as taken from MURPU Engineering Secondary Liquid Mass report from NSSS vendor Intact Steam Generator 0 - 2 hrs: 4.24E+05 lbm 0 - 2 hrs: 4.66E+05 lbm AST added 10% margin to steam releases for Steam Release 2 - 8 hrs: 9.6E+05 lbm 2 - 8 hrs: 1.06E+06 lbm conservatism.

8 - 20 hrs: 9.6E+05 lbm 8 - 24 hrs: 2.11E+06 lbm Primary-Secondary 0.65 gpm to three intact SGs 0.65 gpm to two intact SGs No change Leak Rate 0.35 gpm to faulted SG 0.35 gpm to faulted SG Density Used for 62.4 lbm/ft3 62.4 lbm/ft3 No change Leakage Volume-to-Mass Conversion Duration of Intact SG Not modeled- Not modeled- Tube uncovery does not occur with intact SGs Tube Uncovery After Reactor Trip Time to Cool RCS to 20 hrs 20 hrs No change RHR Cut-in A11-4 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Table 4: MSLB Accident Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Intact Steam Generator 100 100 RG 1.183 Appendix E Section 5.5.4 allows an Iodine partition factor iodine partition factor of 100 for the intact SG.

Intact Steam Generator Not modeled-Alkali Metals not 0.32% (Alkali Metal Carryover is provided for per RG 1.183 Appendix Moisture Carryover considered in CLB MSLB source Partition Factor =312) E Section 5.5.4. Moisture carryover is as Fraction term. specified by NSSS vendor report for MURPU.

Table 5: SGTR Accident Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Maximum Pre-Accident 60 Ci/gm Dose Equivalent I-131 60 Ci/gm Dose Equivalent No change Iodine Spike I-131 Concentration Concurrent Iodine 500 X Equilibrium 335 X Equilibrium RG 1.183 Appendix F Section 2.2 allows the 335 Spike Appearance Rate factor.

Initial Steam Generator 0.1 Ci/gm Dose Equivalent I- 0.1 Ci/gm Dose No change Iodine Source Term 131 Equivalent I-131 Iodine Chemical Form 100% Elemental 0% aerosol, 97% Iodine chemical form is per RG 1.183 Appendix F elemental, 3% organic Section 4.

Percentage of Fuel 0% 0% No change. Note- the SGTR does not result in Rods Failed failed fuel. This is a leaking fuel pre-condition included for conservatism.

RCS Mass 2.53E+08 g 2.258E+08 g AST value as supplied by NSSS vendor.

Steam Generator 4.2E+07 g (each) 4.44 E+07 g (each) AST uses value as specified by NSSS vendor Secondary Liquid Mass MURPU engineering report.

Intact Steam Generator 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months -696500 lbm 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months -696500 lbm No change Steam Release 2 - 20 hr - 2495600 lbm 2 - 20 hr - 2495600 lbm Ruptured Steam 0 - 2 hr - 214000 lbm 0 - 2 hr - 214000 lbm No change Generator Steam 2 - 20 hr - 37100 lbm 2 - 20 hr - 37100 lbm Release Time of Reactor Trip 49.7 s 49.7 s No change Primary-Secondary 1 gpm 1 gpm No change Leak Rate Density Used for 62.4 lbm/ft3 62.4 lbm/ft3 No change Leakage Volume-to-Mass Conversion Ruptured Tube Break 0 - 5502 s - 204900 lbm 0 - 5502 s - 204900 lbm No change Flow Break Flow Flashing Variable 0.15 Constant Constant value for AST is conservative compared Fraction to the variable flow depicted in FSAR Figure 15.6.3-13 Duration of Intact SG 0 minutes 0 minutes No change Tube Uncovery After Reactor Trip Time to Cool RCS to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months No change RHR cut in Intact Steam Generator 100 100 No change Iodine Partition Coefficient Intact Steam Generator CLB source term did not model 0.32% (Partition factor for Carryover is provided for per RG 1.183 Appendix Moisture Carryover Alkali Metals Alkali Metals = 312) F Section 5.6 and Appendix E Section 5.5.4.

Fraction Moisture carryover is as specified by NSSS vendor MURPU engineering report.

A11-5 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Table 6: Control Rod Ejection Accident Inputs and Assumptions Input/Assumption CLB Value for New AST Value for Reason for Change Offsite and Offsite Control Room and Control Room Fuel Rod Gap Fractions Iodine - 12% Iodine/noble gases - 0.10 AST gap fractions are per RG 1.183 Appendix H Kr85 - 10% Other halogens - 0.05 (iodines and noble gases) and Table 3 (other Alkali metals - 0.12 halogens and alkali metals)

Fuel Rod Peaking Factor Not provided 1.7 Radial peaking factor is applied per RG 1.183 Section 3.1.

Percentage of Fuel Rods 10% 10% No change Damaged Percentage of Fuel That 0.25% 0.25% No change Experiences Melting Number of rods exceeding 0 0 No change 6.3 kw/ft above 54 GWD/MTU Initial RCS Iodine Source Term 60 Ci/gm DE I-131 1.0 Ci/gm DE I-131 RG 1.183 Appendix H does not require a preaccident spike and is silent on initial RCS source term. 1.0 Ci/gm is the maximum equilibrium concentration. Initial RCS concentration is small in relation to fuel release.

Initial RCS non-Iodine Activity 1% defective fuel 1% defective fuel (alkali Initial alkali metals based on 1% defective fuel are metals and other halogens) included in AST. Noble gas based on 280 Ci/gm 280 Ci/gm DE Xe-133 DE Xe-133 per TSTF-490.

(noble gases)

Initial Steam Generator 0.1 Ci/gm DE I-131 0.1 Ci/gm DE I-131 No change Iodine Source Term Iodine Chemical Form - Not provided 97% elemental, 3% organic Iodine chemical form is in accordance with RG Secondary Release 1.183 Appendix H Section 5 Iodine Chemical Form - Not provided 95% aerosol, 4.85% Iodine chemical form is in accordance with RG Containment Release elemental, 0.15% organic 1.183 Appendix H Section 4 Containment Volume 2.93E6 ft3 2.93E6 ft3 No change Containment Leakage Rate 5% additional margin added to AST analysis for 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.2% 0.201% conservatism 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days 0.1% 0.105%

Containment Leakage 0% 0% No change Filtration Natural Deposition in 50% plateout of RCS release Elemental iodine - None Natural deposition is credited per RG 1.183 Containment Aerosols - 3.005E-2 hr-1 Appendix H Section 6.1.

Iodine/Particulate Not provided Not credited No change Removal by Containment Sprays RCS Mass 2.3E+08 grams 2.258E+08 grams AST value as supplied by NSSS vendor.

Steam Generator 1.8E+08 grams (4 SGs) 1.778E+08 grams AST uses value as specified by NSSS vendor Secondary Liquid Mass MURPU engineering report.

Primary-Secondary Leak 1 gpm total 1 gpm total No change Rate Duration of Primary-to- 214 seconds 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months Until shutdown cooling established per RG 1.183 Secondary Leakage Appendix H Section 7.1.

Density Used for Leakage 62.4 lbm/ft3 62.4 lbm/ft3 No change Volume-to-Mass Conversion Secondary Steam 49,000 lbm (total) 0-2 hrs - 5.5E+05 lbm Steam releases are assumed until shutdown Release 2-8 hrs - 1.365E+06 lbm cooling established per RG 1.183 Appendix H 8-20 hrs - 2.73E+06 lbm Section 7.1 Duration of SG Tube 0 minutes 0 minutes No change Uncovery Following Reactor Trip Steam Generator Iodine 100 100 No change Partition Coefficient A11-6 1 to Enclosure VEGP AST Accident Analysis Input Values Comparison Tables Table 6: Control Rod Ejection Accident Inputs and Assumptions Input/Assumption CLB Value for New AST Value for Reason for Change Offsite and Offsite Control Room and Control Room Steam Generator Not provided 0.32% (Partition factor for Carryover is provided for per RG 1.183 Appendix Moisture Carryover Alkali Metals = 312) H Section 7.4 and Appendix E Section 5.5.4.

Fraction Moisture carryover is as specified by NSSS vendor MURPU engineering report.

Table 7: Locked Rotor Accident Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Fuel Rod Gap Fractions 10% for all except: I-131 - 0.08 AST gap fractions are per RG 1.183 Table 3 Kr-85, I-127, I-129 = 30% Kr 0.10 Other Halogens and Noble Gases - 0.05 Alkali Metals - 0.12 5% Failed Fuel Fuel Rod Peaking Factor Not Specified 1.7 Radial peaking factor is applied per RG 1.183 Section 3.1.

Number of rods 0 0 Per reload requirements assemblies that could exceeding 6.3 kw/ft exceed 6.3 kW/ft are not loaded in locations above 54 GWD/MTU where DNB could occur.

Initial Steam Generator 0.1 Ci/gm 0.1 Ci/gm No change Iodine Source Term Iodine Chemical Form 100% Elemental 95% particulate Iodine chemical form is in accordance with RG 4.85% elemental 1.183 Appendix G Section 5.6.

0.15% organic RCS Mass 2.3 E+08 g 2.258E+08 lbm RCS mass as supplied by NSSS Vendor Steam Generator 1.9 E+08 g 1.78E+08 lbm AST uses value as specified by NSSS vendor Secondary Liquid Mass MURPU engineering report.

Primary-Secondary Leak 1 gpm 1 gpm total AST value increased for additional conservatism.

Rate Density Used for Not Specified 62.4 lbm/ft3 No change Leakage Volume-to-Mass Conversion Secondary Steam 0 - 2 hr 555000 lbm 0 - 2 hr 610500 lbm AST assumes 10% adder on steam flows for Release 2 - 8 hr 1365000 lbm 2 - 8 hr 1501500 lbm conservatism.

8 - 20 hr 2730000 lbm 8 - 20 hr 3003000 lbm Duration of SG Tube 0 minutes 0 minutes No change Uncovery Following Reactor Trip Steam Generator Iodine 100 100 RG 1.183 Appendix G Section 5.6 allows an Partition Coefficient iodine partition factor of 100 for SG releases.

Intact Steam Generator CLB does not consider Alkali 0.32% (partition Carryover is provided for per RG 1.183 Appendix Moisture Carryover Metals in source term coefficient of 312 for Alkali G Section 5.6 and Appendix E Section 5.5.4.

Fraction Metals) Moisture Carryover is as specified in NSSS vendor report for MURPU A11-7 2 to Enclosure List of Regulatory Commitments Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490 Attachment 12 List of Regulatory Commitments A12-1 2 to Enclosure List of Regulatory Commitments The following table identifies the regulatory commitment in this Attachment to the Enclosure.

REGULATORY COMMITMENT TYPE SCHEDULED COMPLETION DATE/EVENT One Continuing Time Compliance

1. The following guidelines are included in X Prior to the assessment of systems removed implementation from service during movement of of the license irradiated fuel: amendment Ventilation system and radiation monitor availability should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the proposed license amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

A single normal or contingency method to promptly close containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure; rather the prompt methods should enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

A12-2

v t e Letter Sequence Request
EPID:L-2022-LLA-0097, TSTF-286, TSTF-471, TSTF-490, TSTF-051, TSTF-571, TSTF-448 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, ... further results\|Request]] Acceptance... Supplement, Supplement Administration Meeting Questions 1 August 2022 Answers 13 September 2022 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: ML22228A220, ML23024A110, ML23059A456
MONTHYEAR2021-11-05 [Table View]

NL-22-0208, License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490

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NL-22-0208, License Amendment Request for Alternative Source Term, TSTF-51, TSTF-471, and TSTF-490

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