Text

Proposed License Amendment Requests to Update Main Steam Line Break Alternative Source Term Dose Consequences Analyses
	ML25136A297
Person / Time
Site:	North Anna, Surry
Issue date:	05/16/2025
From:	James Holloway; Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	25-017
	Download: ML25136A297 (1)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RIC HMOND, VIRGINIA 23261 May 16, 2025 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Serial No.:

NRA/DPJ:

Docket Nos.:

10 CFR 50.90 25-017 License Nos.:

RO 50-338/339 50-280/281 NPF-4/7 DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUESTS TO UPDATE MAIN STEAM LINE BREAK ALTERNATE SOURCE TERM DOSE CONSEQUENCES ANALYSES Pursuant to 10 CFR 50.90, Virginia Electric and Power Company {Dominion Energy Virginia) requests amendments to North Anna Power Station (NAPS) Units 1 and 2 Subsequent Renewed Facility Operating License Numbers NPF-4 and NPF-7, and Surry Power Station {SPS} Units 1 and 2 Subsequent Renewed Facility Operating License Numbers DPR-32 and DPR-37, respectively.

The proposed License Amendment Requests (LARs) update the Main Steam Line Break (MSLB} Alternate Source Term (AST) dose consequence analysis for NAPS and SPS Units 1 and 2, respectively.

Specifically, the MSLB AST dose consequence analysis for each station has been revised to address extended cooldown timelines that could result from a stagnant Reactor Coolant System loop following a MSLB. Evaluating the MSLB AST dose consequence analysis per 10 CFR 50.59 determined that the activity results in a departure from a method of evaluation used in the safety analysis described in the Final Safety Analysis Report, and would therefore require prior NRG approval. A description and assessment of the proposed change is provided in Enclosure 1 for NAPS Units 1 and 2 and Enclosure 2 for SPS Units 1 and 2.

Dominion Energy Virginia has evaluated the proposed amendments and has determined they do not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in the enclosure for each station. Dominion Energy has also determined operation with the proposed amendments will not result in a significant increase in the amount of effluents that may be released offsite or a significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendments are eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22{c)(9).

Pursuant to 10 CFR 51.22(b}, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed amendments.

Serial No.25-017 Docket Nos. 50-338/339/280/281 Page 2 of 3 The LARs have been reviewed and approved by the respective station's Facility Safety Review Committee.

Dominion Energy Virginia requests approval of the proposed amendments by May 31, 2026, with a 90-day implementation period.

Should you have any questions or require additional information, please contact Mr. Daniel Johnson at (804) 273-2381.

Respectfully, 1 ( (]/)

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Ja~ Holloway Vice President - Nuclear Engineeri g & Fleet Support Commitments contained in this letter: None

Enclosures:

1. License Amendment Request - Updated Main Steam Line Break Alternate Source Term Dose Consequence Analysis - North Anna Power Station Units 1 and 2

2. License Amendment Request - Updated Main Steam Line Break Alternate Source Term Dose Consequence Analysis - Surry Power Station Units 1 and 2 COMMONWEAL TH OF VIRGINIA

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COUNTY OF HENRICO

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The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. James E. Holloway, who is Vice President - Nuclear Engineering & Fleet Support, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this I 0 -+.... day of (Y)c.\\f My Commission Expires: cct--go--zoz-=,-

JULIE H HOUGH NOTARY PUBLIC 8066994 COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES 09-30-2027 I 2025.

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Serial No.25-017 Docket Nos. 50-338/339/280/281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street, Suite 730 Richmond, VA 23219 Mr. L. John Klos NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager - North Anna U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station

Serial No.25-017 Docket Nos. 50-338/339 LICENSE AMENDMENT REQUEST UPDATED MAIN STEAM LINE BREAK ALTERNATE SOURCE TERM DOSE CONSEQUENCE ANALYSIS Virginia Electric and Power Company (Dominion Energy Virginia)

North Anna Power Station Units 1 and 2

Serial No.25-017 Docket Nos. 50-338/339 Page 1 of 40 LICENSE AMENDMENT REQUEST UPDATED MAIN STEAM LINE BREAK ALTERNATE SOURCE TERM DOSE CONSEQUENCE ANALYSIS NORTH ANNA POWER STATION UNITS 1 AND 2

1.0 INTRODUCTION AND BACKGROUND

In 2018, the Callaway Nuclear Plant issued License Event Report (LER) 2018-002-00

[Reference 10], which identified that following a Main Steam Line Break (MSLB) with a loss of offsite power (LOOP), an asymmetric natural circulation cooldown (ANCC) could result in Reactor Coolant System (RCS) flow stagnation in the affected loop. Flow stagnation would then have the effect of extending the timeline to cooldown and depressurize the RCS following a MSLB.

Dominion Energy evaluated the effect of extending the timeline to cooldown and depressurize the RCS following a MSLB and determined that the dose consequence analyses for the MSLB Event at North Anna Power Station (NAPS) was affected. As a result, a new bounding timeline was determined, and new dose consequence analyses were performed. The following sections describe the stagnant loop phenomenon, as well as Dominion Energys bases for maintaining safe shutdown (i.e., MODE 3 - Hot Standby definition in the Technical Specifications (TS)),

10-day steam release duration, and radiological release termination.

The proposed License Amendment Request (LAR) scope is associated with only the MSLB Dose Consequence analysis and does not include the effects (if any) of a stagnant loop condition on other Design Basis Accidents. For other scenarios (e.g., Steam Generator Tube Rupture),

Dominion Energy continues to investigate the effects and will submit separate LARs if required.

Stagnant Loop Phenomenon During a MSLB event, the primary mitigating actions are to isolate feedwater (FW) and Auxiliary Feedwater (AFW), as well as isolate the steaming flow paths for the faulted steam generator (SG). By isolating the SG, the affected RCS loop becomes inactive. Once the SG is isolated and the RCS is stabilized, the unit has reached a safe shutdown condition. After the establishment of safe shutdown, the following phase is to perform an RCS cooldown and depressurization for long-term recovery.

When natural circulation is established with at least one inactive loop, it is known as asymmetric natural circulation (ANC). When ANC occurs, flow through the inactive loop may cease, resulting in a stagnant loop condition. When performing an ANC cooldown with a stagnant loop, the operators control the cooldown of the bulk RCS fluid, but the stagnant loop remains hot since heat transfer in that loop has ceased. Unless flow is re-established permanently or temporarily, the affected RCS loop will remain at elevated temperatures for an extended period of time. This becomes an issue when RCS depressurization to Residual Heat Removal (RHR) system entry

Serial No.25-017 Docket Nos. 50-338/339 Page 2 of 40 conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization.

In this condition, the time required to complete a cooldown to Cold Shutdown (CSD) conditions (i.e., MODE 5 - Cold Shutdown definition in the TS) could be sufficiently extended beyond the current safety analyses assumptions (i.e., 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for NAPS). When evaluating the effects of the stagnant loop condition, as well as potential operational strategies to expedite the cooldown, Dominion Energy determined an extended qualitative bounding timeline was more appropriate.

The bounding timeline allows Operations to establish a safe shutdown condition and then perform either an expedited ANCC or hold at safe shutdown conditions for up to 7 days. Using this assumption, the unit may be held at safe shutdown until either offsite power is restored (allowing start of a Reactor Coolant Pump (RCP) which eliminates the stagnant loop concern),

the faulted SG is recovered (allowing a symmetric plant cooldown which eliminates the stagnant loop concern), or a planned ANCC is performed.

The timeline developed separates expedited ANCC strategies from the dose consequence analyses (i.e., the expedited strategies are not credited in the analysis). By analyzing such a large timeline for steam release, the Operations team and Emergency Response Organization (ERO) will have ample time to recover equipment and reduce unnecessary expedited strategies.

This approach of remaining at safe shutdown (i.e., Hot Standby or MODE 3) conditions for a 10-day period (i.e., 7-day period represents LOOP duration time with a 3-day forced cooldown to RHR entry conditions and RHR cooldown to CSD (i.e., MODE 5)) was used to analyze the dose consequences of an extended MSLB event. It is important to note that the analyses do not credit or require restoration of offsite power as the primary success path for the safety analyses, instead it is chosen as a bounding assumption which envelopes all scenarios that result in a cooldown in less than 10 days.

Basis for maintaining MODE 3 operation:

NAPS is designed and licensed to establish and maintain a safe shutdown condition following a design-basis accident. More explicitly, NAPS is designed and licensed to maintain MODE 3 operation as a safe shutdown mode using the SGs and atmospheric dump valves (SG Power Operated Relief Valves (PORVs)) to remove decay and sensible heat if the condenser steam dump valves are not available. (Note the SG PORVs steam release function is not credited in the safety analyses for overpressure mitigation; therefore, the intact SGs (ISGs) will limit and control pressure utilizing the Main Steam Safety Valves (MSSVs)) [Reference 18]. A subsequent cooldown can be performed using both the ISGs and non-credited or non-safety grade equipment (i.e., SG PORVs and RHR system) when available to complete the cooldown to less than 200°F. It is important to note that since NAPS is a Hot Standby (MODE 3) plant by design and licensing basis, it does not have single-failure proof safety related equipment installed to reach CSD (MODE 5). Therefore, to comply with Regulatory Guide 1.183, Revision 0 [Reference 1], the dose consequence analysis must model steam releases until the unit reaches CSD (i.e.,

MODE 5), and credit must be taken for the use of non-safety related equipment. It is also known that alternative but sufficient water supplies may be needed to support AFW system operation if the plant is required to stay in a safe shutdown condition for an extended period.

Serial No.25-017 Docket Nos. 50-338/339 Page 3 of 40 In summary, the updated analyses model NAPS maintaining safe shutdown in the Hot Standby (i.e., MODE 3) condition for an extended period with a subsequent cooldown to CSD (i.e., MODE

5) conditions when appropriate equipment to perform the cooldown becomes available and when directed by senior station operations personnel. The subsequent cooldown may credit non-safety grade structures, systems and components (SSCs) to achieve CSD mode of operation (i.e., MODE 5).

Basis for 10-day radiological release duration:

A review of industry operating experience (OE) shows that longer full LOOP recovery durations are typically caused by widespread damage due to severe weather events such as tornadoes, hurricanes, and earthquakes. Nuclear plants were designed to withstand natural events based on the historical and environmental data available at the time. However, the industry now has more than 40-50 years of operating history that should also be considered when evaluating LOOP duration times.

The OE shows that for the vast majority of LOOP events, offsite power was restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from initiation. Comprehensive LOOP studies performed by Idaho National Laboratory (INL) and Institute of Nuclear Power Operations (INPO) draw similar conclusions [References 11 - 14]. The studies concluded the vast majority of events are caused by equipment and human performance related issues and are recovered in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . More extensive damage and longer recovery times were more often due to large scale severe weather events. The studies also concluded that LOOP recovery time trended upward due to additional time being taken for investigation prior to restoring the preferred power source [Reference 15].

The longest recorded LOOP recovery time was at Turkey Point in 1992, which took approximately 6 1/2 days to recover offsite power following Category 5 Hurricane Andrew

[References 16 & 17]. In 2017, Category 4 Hurricane Irma struck southern Florida; however, grid power to Turkey Point was never lost. This was due in part to improvements in alternating current (AC) power reliability made on and offsite. Likewise, Dominion Energys Virginia Nuclear Fleet has improved their switchyard design based on this and other relevant OE in recent years, thereby increasing the AC reliability from when the plant was first built and licensed. Probabilistic Risk Assessment (PRA) analysis for NAPS determined the mean recovery time for all LOOP events that occurred during power operation is 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and 21 minutes, while the mean recovery time specifically for weather-related LOOP events is 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months and 11 minutes [Reference 19].

Based on the longest LOOP recovery time being 6 1/2 days during the last 39 years of nuclear industry data, a 7-day LOOP recovery time is considered reasonable.

Note the licensing bases for NAPS does not require the MSLB event be considered coincident with a severe weather event. Therefore, the timeline for expected recovery of offsite power does not need to consider severe weather events. For the purposes of conservatism, the timeline used by Dominion Energy was extended to exceed even those of severe weather-related LOOP durations.

Based on industry weather related LOOP event recovery times, NAPS redundancy of offsite

Serial No.25-017 Docket Nos. 50-338/339 Page 4 of 40 sources, General Design Criteria (GDC) 17 independence of electrical design, and Dominion Energy's demonstrated emergency preparedness and restoration capabilities, a 7-day LOOP recovery time for determining radiological consequences in a safe shutdown condition is reasonable. Therefore, in absence of a regulatory or industry definition, a reasonable LOOP recovery time for determining radiological consequences while at a safe shutdown condition was selected for NAPS. At this point, the recovery of offsite power or recovery of the faulted SG will occur to allow the operators to perform a forced or symmetric cooldown to CSD (i.e., MODE 5) conditions which eliminates the stagnant loop and ANCC concerns.

Once an offsite feeder is restored, an RCP can be started allowing a forced flow cooldown and/or the faulted SG is recovered to allow a symmetric cooldown to RHR entry conditions (assumed 1 day duration), followed by an RHR cooldown to CSD (i.e., MODE 5) (assumed 1 day duration),

and cooldown of the ISGs to < 212 °F (assumed 1 day duration). Therefore, the dose consequences for a MSLB with a LOOP has the potential to last for 10 days, avoiding the stagnant loop and ANCC complexity. Its important to note that specific timing of each phase is irrelevant to the analysis; the total duration of 10 days is the dominating contributor.

Basis for radiological release termination:

Dominion Energy adopted and incorporated the AST methodology for dose consequences at NAPS in 2005 [Reference 6]. Reference 1 Table 6 states that the analysis release duration for a PWR MSLB is until CSD is established (i.e., RCS average coolant temperature 200°F).

Appendix E of Reference 1 states the following under Transport 5.3:

The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Dominion Energys position for the MSLB dose analysis is to assume the steam release rate that was calculated at four hours is extended to 10 days, at which point the release is considered terminated. Using this conservative 10-day approach on an MSLB event to CSD (i.e., MODE 5) provides a bounding dose consequence, which allows the operators to select the desired path based on actual plant parameters, equipment availability, and operations leadership direction to achieve CSD (i.e., MODE 5) (termination of the radiological consequence). Increasing the extended cooldown timeline to avoid an ANCC situation would result in increased dose consequences that exceed the more than minimal threshold for 10 CFR 50.59.

Figure 1-1 demonstrates a simplified graphical representation of the three distinct phases considered for the MSLB dose analysis timeline. This timeline is based on the discussions above for each phase.

Serial No.25-017 Docket Nos. 50-338/339 Page 5 of 40 Figure 1-1: Phases of MSLB Accident Mitigation 2.0 RADIOLOGICAL CONSEQUENCES INTRODUCTION This report describes the evaluations conducted to assess offsite doses and Control Room habitability at North Anna Power Station (NAPS) following a postulated MSLB design basis accident per Regulatory Guide (RG) 1.183, Revision 0 [Reference 1]. The accident source term discussed in Reference 1 is herein referred to as the Alternative Source Term (AST). The evaluations have employed the detailed methodology contained in RG 1.183 for use in design basis accident analyses for the AST. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 [Reference 2] or the supplemental guidance in RG 1.183.

This application includes the following key elements:

Increase the time required for cooldown; Adding an MSSV failure scenario; Include new control room atmospheric dispersion factors; Credit the use of control room emergency ventilation; Modifying the secondary side liquid source term to be 10% of the primary coolant source term, excluding noble gases (NGs).

3.0 CURRENT LICENSING BASIS

SUMMARY

The current MSLB design basis radiological analysis which appears in NAPS Updated Final Safety Analysis Report (UFSAR) Section 15.4.2 was submitted for approval in Reference [5] and approved in References [6] and [7]. The analysis was performed using the RADTRAD-NAI code based on a core inventory derived with the ORIGEN-S code.

Event Mitigation

Use of only Safety Related (SR)

Equipment that is protected by Technical Specifications

MSLB Isolation

RCS Stablization at Safe Shutdown Conditions Hold at Safe Shutdown

Long-Term Recovery

Follows Transient Mitigation

May result in makeup to SR water sources from non-safety (NS) sources

May be up to 7 days Cooldown /

Depressurization

Long-Term Recovery

Cooldown to CSD IAW Dose Consequence RG Requirements

Uses NS Equipment

May take up to 3 additional days

Serial No.25-017 Docket Nos. 50-338/339 Page 6 of 40 4.0 ANALYSIS ASSUMPTIONS AND KEY PARAMETER VALUES This section describes the general analysis approach and presents analysis parameters, Table 4-1, and assumptions, Table 4-2. The analysis parameters and assumptions presented in Table 4-1 and Table 4-2 have not changed from the values documented in References [5] and [6].

The dose analyses documented in this application employ the Total Effective Dose Equivalent (TEDE) calculation method, as specified in Reference [1] for AST applications. The TEDE is determined at the Exclusion Area Boundary (EAB) for the worst 2-hour interval. TEDE values for individuals at the Low Population Zone (LPZ) and for NAPS Control Room personnel are calculated for the assumed duration of the event.

The TEDE concept is defined to be the Deep Dose Equivalent, DDE (from external exposure) plus the Committed Effective Dose Equivalent, CEDE (from internal exposure). In this manner, TEDE assesses the impact of relevant nuclides upon body organs, in contrast with the previous single, critical organ (thyroid) concept for assessing internal exposure. CEDE dose conversion factors were taken from Table 2.1 of Federal Guidance Report (FGR) 11, Reference [8], per Section 4.1.2 of Reference [1]. The DDE is nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE is used instead of DDE in determining the contribution of external dose to the TEDE. EDE dose conversion factors were taken from Table III.1 of FGR 12, Reference [9], per Section 4.1.4 of Reference [1].

Table 4-1: Analysis Parameters Parameter Value Control Room Normal Intake Unfiltered Flow 2800 cfm Control Room Volume 230,000 ft3 Turbine Building Free Air Volume 6E6 ft3 RCS total leak rate to 3 Steam Generators 1 gallon/min Moisture Carryover in Intact SG 1%

Iodine Partitioning Source Fractional Value ASG 1.0 ISG 0.01 EAB Atmospheric Dispersion Factor Time (hr)

Dispersion Factor (s/m3) 0 - 720 3.10E-04 LPZ Atmospheric Dispersion Factor Time (hr)

Dispersion Factor (s/m3) 0 - 2 1.10E-05 2 - 8 1.10E-05 8 - 24 7.30E-06

Serial No.25-017 Docket Nos. 50-338/339 Page 7 of 40 Table 4-1: Analysis Parameters Parameter Value 24 - 96 3.00E-06 96 - 720 8.20E-07 Control Room Breathing Rate Time (hr)

Breathing Rate (m3/s) 0 - 720 3.5E-04 LPZ Breathing Rate Time (hr)

Breathing Rate (m3/s) 0 - 8 3.5E-04 8 - 24 1.8E-04 24 - 720 2.3E-04 Iodine Chemical Form Product Fractional Distribution Elemental Iodine 0.97 Methyl Iodine (organic) 0.03 Aerosol (particulate) 0.0 RCS TS Activity Limit for Dose Equivalent I-131 (TS SR 3.4.16.2) 1.0 µCi/gm RCS TS Activity Limit for Dose Equivalent Xe-133 (TS SR 3.4.16.1) 197 µCi/gm Secondary Specific Activity TS Limit for Dose Equivalent I-131 (TS SR 3.7.7.1) 0.1 µCi/gm Table 4-2: Analysis Assumptions Assumption Description Turbine Building Exhaust Rate 0.2 volumes/hr (Note 1)

Note 1: This assumption does not affect the analyses associated with the faulted MSSV. The case modeling a break in the Turbine Building uses the low turnover rate to maximize the Control Room dose.

Serial No.25-017 Docket Nos. 50-338/339 Page 8 of 40 4.1 Proposed Licensing Basis Changes This section provides a summary description of the key proposed licensing basis changes that are justified with the revised NAPS AST MSLB analysis. This LAR is being submitted for prior NRC review and approval pursuant to the requirements of 10 CFR 50.59 which specifies that a departure from a method described in the UFSAR, such as the design basis radiological consequence analyses, shall be submitted for approval unless the changes to the elements of the method meet certain requirements.

The proposed changes for a MSLB radiological event are:

1) An increase of the cooldown times associated with LOOP conditions,

2) Adding a malfunctioning MSSV as a release path (due to extended release time and potential for new limiting scenario),

3) Crediting the use of control room emergency ventilation,

4) Including new control room atmospheric dispersion factors, and

5) Modifying the secondary side liquid source term to be 10% of the primary coolant source term, excluding noble gases.

The proposed changes have been analyzed and result in acceptable consequences, thereby meeting the criteria as specified in References [1] and [2]; however, they did not meet the requirements for implementation under 10 CFR 50.59 without prior NRC approval.

Serial No.25-017 Docket Nos. 50-338/339 Page 9 of 40 Table 4-3: Comparison of Proposed Changes Parameter CLB Value Proposed Value Reason for Change SG Liquid Volume/Mass per SG 2,054 ft3 109,160 lbm Hot Full Power (HFP) conditions, consistent with the source document for steam flows RCS Liquid Volume/Mass 9,786 ft3 404,459 lbm HFP conditions, consistent with the source document for steam flows Control Room Unfiltered Inleakage 500 cfm 250 cfm Conservatively bounds the Control Room in-leakage testing results of 102 +/- 8.9 cfm (Reference 24)

RCS Leakage Liquid Density 47.5054 lbm/ft3 62.4 lbm/ft3 Reflects guidance found in RG 1.183 (Reference 1)

EAB Breathing Rate Time (hrs) Breathing Rate (m3/s)

Time (hrs)

Breathing Rate (m3/s)

Use of conservative breathing rate for entire event 0 - 8 3.5E-04 0 - 720 3.5E-04 8 - 24 1.8E-04 24 - 720 2.3E-04 Control Room Occupancy Factors Time (hrs)

Occupancy Factor Time (hrs)

Occupancy Factor Reflects guidance found in RG 1.183 (Reference 1) 0 - 8 1.0 0 - 8 1.0 8 - 24 1.0 8 - 24 1.0 24 - 96 0.6 24 - 96 0.6 96 - 720 0.6 96 - 720 0.4

Serial No.25-017 Docket Nos. 50-338/339 Page 10 of 40 Table 4-3: Comparison of Proposed Changes Parameter CLB Value Proposed Value Reason for Change Release Duration, hours 8

LOOP 240 No-LOOP 12 Recovery from Stagnant Loop after 10 days Normal CR Intake Atmospheric Dispersion Factor Time (hr)

Dispersion Factor (s/m3)

Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3)

Revised ARCON96 calculations (updated true north) and crediting steam flow velocities. See Section 4.2.4.

0-2 1.04E-02 0-2 1.05E-02 2.10E-03 2-8 8.20E-03 2-8 8.63E-03 1.73E-03 8-24 3.23E-03 8-24 3.44E-03 6.88E-04 24-96 2.25E-03 24-96 2.39E-03 4.78E-04 96-720 1.68E-03 96-720 1.78E-03 3.56E-04 Table 4-4: Additional Analysis Parameters Associated with Crediting Control Room Emergency Ventilation Parameter Proposed Value Basis Control Room Emergency Intake Filtered Flow 900 cfm Minimum value of TS ventilation flow rate of 1000 cfm +/- 10%.

Control Room Isolation, hours (seconds) Turbine Building Releases 0.0167 (60)

Faulted MSSV 0.044444 (160)

The CR ventilation damper closure time is designed to accommodate the Safety Injection (SI) actuation time as well as a 20 second damper delay.

Additional time has been included as retained margin for future use.

Serial No.25-017 Docket Nos. 50-338/339 Page 11 of 40 Table 4-4: Additional Analysis Parameters Associated with Crediting Control Room Emergency Ventilation Parameter Proposed Value Basis Main Control Room Emergency Fan Actuation Time for Filtered Pressurization 60 minutes after the SI (modeled as 60 minutes after the start of a MSLB)

The updated MSLB analysis credits the existing time critical operator action for the loss of coolant accident, fuel handling accident, and the steam generator tube rupture, which is already controlled under Dominions Time Critical Action Program Control Room Filter Efficiencies Product Efficiency (%)

Specific guidance is found in Reg Guide 1.52, as referenced by T.S.

5.5.10. The efficiencies given here include a safety factor of 2.

Elemental Iodine 95 Methyl Iodine (organic) 95 Aerosol (particulate) 98 Emergency CR Intake Atmospheric Dispersion Factor Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3)

See Section 4.2.4 0-2 3.33E-03 6.66E-04 2-8 2.41E-03 4.82E-04 8-24 9.53E-04 1.91E-04 24-96 6.67E-04 1.33E-04 96-720 4.93E-04 9.86E-05

Serial No.25-017 Docket Nos. 50-338/339 Page 12 of 40 4.2 Main Steam Line Break (MSLB) Reanalysis This application involves the reanalysis of the design basis radiological analyses for the MSLB.

The calculated radiological consequences are compared with the acceptance criteria provided in 10 CFR 50.67(b)(2), as clarified per the additional guidance in RG 1.183 for events with a higher probability of occurrence.

Dose calculations are performed at the EAB for the worst 2-hour period, and for the LPZ and NAPS Control Room for the duration of the accident. The radiological dose consequence calculations were performed with the RADTRAD-NAI computer code system [Reference 3]. The applicable dose acceptance criteria are provided in Table 4-5.

Table 4-5: MSLB Accident Dose Acceptance Criteria Accident Control Room (Note 1)

EAB & LPZ Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Concurrent Iodine Spike 5 rem TEDE 2.5 rem TEDE Note 1: Based on 10 CFR 50.67 and 10 CFR 50, Appendix A, GDC 19 [Reference 4]

This analysis includes doses associated with the releases of radioactive material initially present in primary and secondary liquids at maximum allowable TS concentrations plus iodine spiking scenarios. No fuel failure is expected. Doses were calculated at the EAB, LPZ, and Control Room. The methodology used to evaluate the dose consequences resulting from the MSLB is consistent with RG 1.183 [Reference 1].

4.2.1 MSLB Scenario Description The MSLB accident begins with a break in one of the Main Steam lines leading from an affected SG (ASG) to the turbine. To maximize doses, break scenarios are assessed as follows: 1) a break in the Turbine Building, and 2) a faulted MSSV.

4.2.1.1 Turbine Building Release The ASG blows down into the Turbine Building for 30 minutes, after which it is isolated (via an existing time-critical operator action) and releases from this pathway stop. Two scenarios are modeled for the MSLB: Loss of Offsite Power (LOOP) and Offsite Power Available (no-LOOP).

Each scenario uses steam release rates specific to the scenario. In addition, both scenarios model Turbine Building exhaust differently. The LOOP scenario models the Turbine Building exhaust fans as not having the power required to operate. Therefore, natural circulation (0.2 building-volumes per hour) is modeled for the LOOP scenario. The no-LOOP scenario models the Turbine Building exhaust flow rate at the maximum capability of the Turbine Building exhaust fans (10 building-volumes per hour). The total modeled flow rate from the Turbine Building includes the ASG blow down flow rates. To maximize dose consequences, neither scenario credits condenser availability.

Serial No.25-017 Docket Nos. 50-338/339 Page 13 of 40 Primary and secondary activity discharged from the ASG is released directly to the Turbine Building without mitigation. Primary-to-secondary leakage in the ASG is assumed at 500 gpd (2.896 lbm/min); total primary-to-secondary leakage modeled from the SGs is 1440 gpd (1 gpm accident induced leak rate specified in TS 5.5.8.b.2). The TS RCS Operational Leakage primary-to-secondary leak rate is 150 gpd through any one SG as specified in TS 3.4.13.d.

The primary system is cooled down through the release of steam from the two ISGs. Steam releases from the ISGs are modeled to occur through the MSSVs and TDAFW turbine exhaust.

Modeling a release through the MSSVs bounds the use of either the PORVs or condenser steam dumps. The ISGs are modeled to have a total leakage rate equal to 940 gpd, which is the remaining primary-to-secondary leakage (1440 gpd - 500 gpd). ISG steaming will continue until sufficient cooldown allows the RHR system to be placed into service. The times for the RHR system to be placed into service are modeled as 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the initiation of the event for the LOOP and no-LOOP scenarios, respectively.

4.2.1.2 Faulted MSSV Release The ASG vents directly to the atmosphere for the duration of the event. The scenario is modeled with a LOOP. The no-LOOP scenario is non-limiting because SI would not occur; further, with offsite power available a normal cooldown and depressurization would occur. There are no releases to the Turbine Building. To maximize dose consequences, the condenser is not credited as being available.

AFW additions to the ASG may occur for up to 30 minutes. The limiting case was an immediate termination of AFW followed quickly by SG dry-out. Steam flow to the environment was increased by a factor of 10 to ensure all activity in the ASG was quickly released to the environment.

Primary and secondary activity discharged from the ASG is released directly to the environment, modeling SG dry-out. Primary-to-secondary leakage in the ASG is assumed at 500 gpd; total primary-to-secondary leakage modeled from the SGs is 1440 gpd (1 gpm accident induced leak rate specified in TS 5.5.8.b.2). The TS RCS Operational Leakage primary-to-secondary leak rate is 150 gpd through any one SG as specified in TS 3.4.13.d.

Cooldown of the primary system is through the release of steam from the two ISGs. Steam releases from the ISGs are modeled to occur through the MSSVs and TDAFW turbine exhaust.

An actual plant cooldown would require steam release through the SG PORVs or condenser steam dumps; however, the modeling assumption (in the dose consequence analysis) of MSSV is representative for the PORVs and conservative relative to condenser steam dumps. The ISGs are modeled to have a total leakage rate equal to 940 gpd, which is the remaining primary-to-secondary leakage (1440 gpd - 500 gpd). ISG steaming will continue until sufficient cooldown allows the RHR system to be placed into service. The time for the RHR system to be placed into service is modeled as 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months for the LOOP scenario.

4.2.2 MSLB Source Term Definition The primary coolant source term definition is unchanged from the CLB (Reference 5). The analysis of the MSLB accident indicates that no fuel rod failures occur as a result of the transient.

Thus, radioactive material releases during the event are determined by assuming the

Serial No.25-017 Docket Nos. 50-338/339 Page 14 of 40 radionuclide concentrations initially present in primary and secondary liquid are at maximum TS limits plus iodine spiking. In accordance with RG 1.183, Appendix E, two independent cases are evaluated. Case one assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike.

The MSLB analysis uses the primary coolant concentration shown in Table 4-6 and the pre-accident iodine spike source term shown in Table 4-7. Initial secondary side liquid concentration is 10% of the primary concentration (based on the ratio of RCS and SG TS Dose Equivalent (DE) I-131 limits) without noble gases. Initial secondary side steam concentrations are negligible.

The MSLB analysis models the concurrent iodine spike, as shown in Table 4-8, which corresponds to an accident-initiated value 500 times the equilibrium appearance rate persisting for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Table 4-6: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Br-84 1.294E-02 Rb-88 1.127E+00 Rb-89 3.072E-02 I-131 7.527E-01 I-132 2.742E-01 I-133 1.217E+00 I-134 1.697E-01 I-135 6.550E-01 Cs-134 8.056E-02 Cs-136 4.475E-02 Cs-137 4.036E-01 Cs-138 2.844E-01 Kr-85 1.570E+00 Kr-85m 6.469E-01 Kr-87 3.743E-01 Kr-88 1.131E+00 Xe-133 8.666E+01 Xe-133m 9.602E-01 Xe-135 1.880E+00 Xe-135m 5.818E-02 Xe-138 2.067E-01

Serial No.25-017 Docket Nos. 50-338/339 Page 15 of 40 Table 4-7: Pre-accident Iodine Spike - 60 µCi/gm DE I-131 (see Note 1)

Nuclide Concentration

(µCi/gm)

I-131 4.516E+01 I-132 1.645E+01 I-133 7.302E+01 I-134 1.018E+01 I-135 3.930E+01 Note 1: The spike was implemented as shown in Table 4-9. The spike was modeled using a TS iodine source with a release fraction of 1 and an iodine spike source with an iodine release fraction of 59. In this manner the complete pre-accident iodine spike of 60 µCi/gm DE I-131 was modeled.

Table 4-8: Concurrent Iodine Spike MSLB RCS Concentration Nuclide Spike = 500 x Appearance Rate (Ci/hr)

Iodine Released 8-hour Spike Duration (Ci)

I-131 1.155E+04 9.240E+04 I-132 1.231E+04 9.848E+04 I-133 2.226E+04 1.781E+05 I-134 1.583E+04 1.266E+05 I-135 1.659E+04 1.327E+05 The RCS concentrations in Table 4-6 were used to build a nuclide inventory file (NIF) that was used to model the RCS, SG liquid, and pre-accident spike activities. To obtain total radioisotope inventories for each compartment from the RCS 1.0 µCi/gm DE I-131 NIF, the concentrations from the NIF were converted using a power level (within RADTRAD-NAI) derived from a combination of compartment masses, unit conversion factors, and other parameters that affect radioisotope inventories in a compartment. For example, the source term fraction entry for the RCS compartment is determined by multiplying the mass of the RCS (lbm) by conversion factors for gm/lbm and Ci/µCi:

RCS = MRCS (lbm) x 453.59237 (gm/lbm) x 1.00E-06 (Ci/µCi)

Iodine and particulate source term fractions of the initial SG liquid compartments for the ASG and ISG were determined using the mass of the SG liquid in the same manner. However, the SG sources also used a release fraction of 0.1 to reflect the differences in TSs. Additionally, the

Serial No.25-017 Docket Nos. 50-338/339 Page 16 of 40 source term is split between the ASGs and ISGs based on their individual liquid masses.

Table 4-9 contains the parameters necessary to model the source term fractions. The use of the data developed in Table 4-9, in conjunction with the RCS activity NIF, based on Table 4-6, allows the RCS, SG liquid, and the two spike activity releases to be modeled in multiple RADTRAD-NAI cases.

Table 4-9: Source Term Fraction Values by Compartment and Source Compartment / Source Compartment Source Term Fraction RFT Release Fraction Values PSF Power Level (Scaling Factor)

RCS TS Source 1

1 183.5 RCS Noble Gases Only 1

1 183.5 RCS Pre-Incident Iodine Spike 1

59 for Iodine, 0 for others 183.5 RCS Concurrent Iodine Spike 1

1 for Iodine, 0 for others 1

Steam Generator TS Source 0.667 for ISG 0.333 for ASG 0.1, 0 for Noble Gases 148.5 4.2.3 MSLB Release Transport 4.2.3.1 Turbine Building Release For the ASG, the release pathway is assumed to be directly into the Turbine Building with no credit taken for partitioning or scrubbing of the SG liquid. From the Turbine Building, the activity is assumed to pass into the Control Room via emergency (filtered) intakes and unfiltered inleakage as well as pass into the environment through vents, louvers, and other openings located around the Turbine Building. The ASG will release activity into the Turbine Building until isolated at 30 minutes. The release from the Turbine Building is determined by each scenario.

In the LOOP scenario, the Turbine Building ventilation system is not operating due to offsite power being unavailable. In the no-LOOP scenario, the system is determined to be energized and is modeled as operating at its maximum capacity. The models use 0.2 and 10 building-volumes per hour for each scenario, respectively.

The ASG transport model that is utilized for noble gases, iodine, and particulates is consistent with Appendix E of RG 1.183. During the first 30 minutes of the event while the ASG is blowing down, radioactivity in the bulk liquid is released without reduction for partitioning or scrubbing.

The primary-to-secondary leak rate in the ASG is 500 gpd. The primary-to-secondary leak rate path associated with the ASG is direct to the Turbine Building and is terminated at the end of the event.

Serial No.25-017 Docket Nos. 50-338/339 Page 17 of 40 The ISGs discharge to the environment until RHR is placed in service. The primary-to-secondary leak rate total to the ISGs is 940 gpd, which is the remainder of the modeled 1 gpm (1440 gpd) total accident induced leakage. Due to the effects of partitioning and moisture carryover, the total radionuclides released to the environment are reduced by a factor of 100. The effect of partitioning and moisture carryover is modeled by reducing the steam release rate by a factor of 100 to conserve radionuclides in the ISG liquid. Releases of noble gases are modeled without reduction for partitioning or moisture carryover. Radionuclides initially in the steam space do not provide any significant dose contribution and are not considered.

There are several nuclide transport models associated with the ASGs and ISGs. The combined results from the cases ensure proper accounting of iodine, particulates, and noble gas releases.

Those models are:

1. Release of secondary side bulk liquid that has activity at the TS limit of 0.1 Ci/gm DE I-131. It is assumed that Noble gas is not present in the bulk liquid initial inventory.

2. Release of TS levels of RCS noble gases activity associated with primary-to-secondary leakage.

3. Release of TS levels of RCS iodine and particulate activity associated with primary-to-secondary leakage.

4. Release of RCS activity associated with concurrent (500 times the appearance rates generated from RCS activity at 1 Ci/gm DE I-131 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) or pre-accident iodine spiking (60 Ci/gm Dose Equivalent I-131).

4.2.3.2 Faulted MSSV Release The faulted MSSV scenario models the ISGs as described in Section 4.2.3.1. However, the ASG release pathway is not into the Turbine Building, but directly to the environment. The steam releases from the ASG were increased by a factor of 10 to simulate immediate isolation of AFW and subsequent dry-out of the ASG. No iodine partitioning or moisture carryover was modeled in the ASG. Primary-to-secondary leakage for the ASG is assumed to flash directly to steam and released directly to the environment. There are no releases into the Turbine Building. The models used in this scenario are similar to those documented in Section 4.2.3.1.

4.2.4 MSLB Atmospheric Dispersion Factors The unchanged CLB EAB and LPZ Atmospheric Dispersion Factors (X/Q) from References 5 and 6 are used. The EAB location uses the worst 2-hour /Q value for the entire timeline and determines the worst 2-hour dose with a sliding sum.

The CLB Control Room X/Qs were submitted in References 5 and 20. The subsequent Safety Evaluation (Reference 6) found the Normal Control Room Intake X/Q values to be acceptable for use. The Emergency Control Room Intake X/Qs were not reviewed at that time as they are not used in the CLB.

Serial No.25-017 Docket Nos. 50-338/339 Page 18 of 40 The Control Room X/Qs have been re-evaluated to address an improved value of the correction for true north. Attachment 1, Section 1.3, of Reference 20 described the true north correction as 36 degrees. A more recent, updated value of the true north correction was submitted in Reference 21 as 23.207 degrees. The proposed Control Room normal and emergency ventilation X/Qs included in Table 4-3 and Table 4-4 include the true north correction of -23.207 degrees instead of -36 degrees.

The re-evaluated Control Room X/Qs were determined using the ARCON96 code (Reference

22) and guidance from Regulatory Guide 1.194 (Reference 23).

4.2.4.1 Meteorological Data The meteorological data used in the Control Room X/Q re-evaluation are unchanged from the CLB (Reference 20) and the recent submittal for the Technical Support Center (TSC) relocation (Reference 21).

4.2.4.2 Dimensional Data The sources included in the re-evaluation were limited to the Unit 1 and Unit 2 PORVs. These sources are applicable to the MSLB. X/Q values calculated for the PORVs are used for the MSSVs and TDAFW turbine exhaust given the relative locations of the release points. The MSSVs are located between the respective containment walls and the PORVs, i.e., slightly further from the control room intakes than the PORVs. The TDAFW turbine exhausts are located in the Auxiliary Feedwater House, which is located past the RWST and further from the control room than the PORVs. The North Anna meteorological tower, receptor, and source locations and elevations remain unchanged from the CLB (Reference 20). The applicable sources and receptors, shown on Sketch No. 1 of Reference 20, are:

P1 - Unit 1 PORVs

P2 - Unit 2 PORVs

NCR - normal control room intake

C4 - emergency control room intake close to column lines C and 4

C6 - emergency control room intake close to column lines C and 6

C10 - emergency control room intake close to column lines C and 10

C11 - emergency control room intake close to column lines C and 11 These source receptor pairs are also applicable to the SG tube rupture (SGTR) and locked rotor accident (LRA) dose consequences analyses.

Serial No.25-017 Docket Nos. 50-338/339 Page 19 of 40 4.2.4.3 Directional Data The directional data presented in Section 1.3 of Reference 20 was revised to incorporate the updated true north correction. Table 1 of Reference 20 is reproduced here to summarize the updated ARCON96 inputs. Only the data associated with the Unit 1 and Unit 2 PORVs as sources are included.

Serial No.25-017 Docket Nos. 50-338/339 Page 20 of 40 Table 4-10: ARCON96 Inputs LOCATION (grade elev 271')

Point Sources Elevation (feet)

Meters above grade Feet South of C Feet East of 9 Feet West of 9 Meters to C-4 Meters to C-11 Meters to C-10 Meters to C-6 Meters to norm intake Degrees from NCR intake to Degrees from C-4 to Degrees from C-11 to Degrees from C-10 to Degrees from C-6 to Unit 1 PORV A 339.00 20.73 117.58 129.08 37.66 70.29 57.79 35.76 22.39 205 175 97 104 138 Unit 1 PORV C 339.00 20.73 117.58 140.58 36.69 73.33 60.63 37.01 19.92 198 170 96 102 133 Unit 1 PORV B 339.00 20.73 117.58 153.50 35.98 76.79 63.90 38.76 17.57 188 164 94 100 128 Unit 2 PORV C 339.00 20.73 117.58 153.25 104.38 43.86 52.81 82.09 103.80 239 227 192 206 222 Unit 2 PORV B 339.00 20.73 117.58 140.33 100.69 41.70 49.91 78.52 99.91 238 226 188 203 221 Unit 2 PORV A 339.00 20.73 117.58 128.83 97.42 40.00 47.45 75.37 96.44 238 225 184 200 220

LBDCR No.25-017 Docket Nos. 50-338/339 Page 21 of 40 4.2.4.4 Wind Speed Statistical Data According to Regulatory Guide 1.194 (Reference 23), releases from atmospheric relief valves which release effluent vertically at high velocity without obstruction qualify for a reduction in the atmospheric dispersion factors computed by ARCON96. This reduction by a factor of five (referred to as divide-by-5) is permitted as long as the vertical velocity of the effluent is at least five times the speed of the wind at the elevation of the release.

The CLB includes a minimum release velocity of 5.72 meters per second at the elevation of the release. This value was determined using linear interpolation between the elevations of the upper and lower met tower instruments.

The minimum release velocity has been re-evaluated using a wind speed power law, which takes the form:

where u is the wind speed at height z; ur is the wind speed at a reference height zr, and the exponent is an empirically derived coefficient.

Fitting the wind speed power law to the 95th percentile wind speed data of Appendix 1, Section 1.4 of Reference 20 results in a 95th percentile wind speed of 6.01 m/s. Thus, the minimum release velocity for divide-by-5 is 30.07 meters per second, or 98.7 ft/sec.

4.2.4.5 Applicability of X/Q Reductions Steam velocities were evaluated for both the MSSV and TDAFW release points. The MSSV steam velocities justify use of X/Qs with divide-by-5, with the ISG steam release velocities exceeding 800 ft/sec and the average ASG steam release velocities post-dryout for the faulted-MSSV cases exceeding 100 ft/sec. The orientation of the TDAFW releases preclude the use of X/Qs with divide-by-5.

4.2.4.6 ARCON96 Results The results of the ARCON96 runs are shown in Table 4-11. The results do not include a reduction factor of 5. The limiting control room normal intake X/Qs were determined by taking the largest value for each time interval from all of the point sources. Similarly, the limiting control room emergency intake X/Qs were determined for each time interval by selecting the largest value over the range of point sources and control room intakes (C-4, C-6, C-10, and C-11). The resulting X/Q values used to calculate the limiting dose consequences are shown in Table 4-12.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 22 of 40 Table 4-11: ARCON96 Results Point Sources Time Interval X/Q to C-4 X/Q to C-11 X/Q to C-10 X/Q to C-6 X/Q to NCR intake Unit 1 PORV A 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2.98E-03 1.05E-03 1.52E-03 3.33E-03 7.56E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.41E-03 6.82E-04 1.01E-03 2.06E-03 6.59E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 9.53E-04 2.39E-04 3.41E-04 7.66E-04 2.56E-03 1 to 4 days 6.60E-04 1.85E-04 2.64E-04 5.64E-04 1.85E-03 4 to 30 days 4.93E-04 1.44E-04 2.03E-04 4.20E-04 1.38E-03 Unit 1 PORV C 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.17E-03 9.65E-04 1.39E-03 3.13E-03 9.01E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.39E-03 6.34E-04 9.32E-04 1.94E-03 7.54E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 9.40E-04 2.24E-04 3.16E-04 7.14E-04 3.01E-03 1 to 4 days 6.67E-04 1.72E-04 2.45E-04 5.23E-04 2.10E-03 4 to 30 days 4.91E-04 1.36E-04 1.87E-04 3.85E-04 1.58E-03 Unit 1 PORV B 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.19E-03 8.95E-04 1.27E-03 2.92E-03 1.05E-02 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.35E-03 5.75E-04 8.46E-04 1.80E-03 8.63E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 8.99E-04 2.07E-04 2.88E-04 6.65E-04 3.44E-03 1 to 4 days 6.28E-04 1.57E-04 2.23E-04 4.97E-04 2.39E-03 4 to 30 days 4.75E-04 1.27E-04 1.72E-04 3.50E-04 1.78E-03 Unit 2 PORV C 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.52E-04 2.45E-03 1.88E-03 8.37E-04 5.69E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.57E-04 2.02E-03 1.61E-03 7.30E-04 4.45E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75E-04 8.03E-04 6.36E-04 2.78E-04 1.77E-04 1 to 4 days 1.34E-04 5.67E-04 4.56E-04 2.08E-04 1.34E-04 4 to 30 days 1.02E-04 4.22E-04 3.40E-04 1.58E-04 1.07E-04 Unit 2 PORV B 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.80E-04 2.63E-03 2.08E-03 9.12E-04 6.13E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.93E-04 2.17E-03 1.77E-03 7.90E-04 4.79E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.88E-04 8.68E-04 6.99E-04 3.04E-04 1.88E-04 1 to 4 days 1.43E-04 6.04E-04 4.99E-04 2.26E-04 1.43E-04 4 to 30 days 1.09E-04 4.49E-04 3.73E-04 1.71E-04 1.15E-04 Unit 2 PORV A 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6.27E-04 2.73E-03 2.25E-03 9.92E-04 6.45E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 5.20E-04 2.31E-03 1.93E-03 8.46E-04 5.14E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.03E-04 9.15E-04 7.64E-04 3.30E-04 2.02E-04 1 to 4 days 1.52E-04 6.27E-04 5.42E-04 2.43E-04 1.52E-04 4 to 30 days 1.16E-04 4.71E-04 4.04E-04 1.84E-04 1.21E-04 Table 4-12: Limiting PORV/MSSV/TDAFW Exhaust X/Q Values Time Interval Normal Control Room Ventilation Intake Normal Control Room Ventilation Intake with Divide-By-5 Emergency Control Room Ventilation Intakes Emergency Control Room Ventilation Intakes with Divide-By-5 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.05E-02 2.10E-03 3.33E-03 6.66E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.63E-03 1.73E-03 2.41E-03 4.82E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.44E-03 6.88E-04 9.53E-04 1.91E-04 1 to 4 days 2.39E-03 4.78E-04 6.67E-04 1.33E-04 4 to 30 days 1.78E-03 3.56E-04 4.93E-04 9.86E-05

LBDCR No.25-017 Docket Nos. 50-338/339 Page 23 of 40 The X/Q values with divide-by-5 presented in Table 4-12 are also applicable to the SGTR and LRA dose consequence analyses based on the discussion in References 5 and 20.

However, the emergency control room X/Qs only apply after a SI signal, which is not certain for the LRA as described in Section 3.5.5 of Reference 5. Therefore, it is intended that the Normal and Emergency Control Room Intake with divide-by-5 X/Q values will be used in future SGTR analyses, and the Normal Control Room Intake with divide-by-5 X/Q values will be used in future LRA analyses as approved changes in elements of a method in accordance with 10 CFR 50.59. The dose consequences of the SGTR and LRA are not submitted with this change because use of the new Normal Control Room X/Qs would result in dose consequences that are essentially the same as those documented in the UFSAR, and the use of Emergency Control Room X/Q values for a SGTR would result in dose consequences significantly lower than those documented in the UFSAR.

4.2.5 Key Analysis Assumptions and Inputs 4.2.5.1 Method of Analysis The RADTRAD-NAI code [Reference 3] is used to calculate the radiological consequences from airborne releases resulting from a MSLB at NAPS to the EAB, LPZ, and Control Room.

The schematic shown in Figure 4-1 provides an overall picture of the design basis MSLB involving a break into the Turbine Building and releases to the environment. Hot full-power (HFP) mass values were used for secondary side bulk liquid masses consistent with the modeled steam releases.

The evaluation of the break in the Turbine Building considered assumptions which maximize dose to model the release from the ASG into the Turbine Building volume. High and low escape rates from the louvers, dampers, etc. (i.e., 10 building-volumes/hr down to 0.2 building-volume/hr) are modeled in separate scenarios to maximize resulting offsite and Control Room dose consequences, respectively.

Releases from the ASG persist for 30 minutes after event initiation, at which point the steam line break is isolated. Releases from the ISGs persist until the RHR system is placed into service. The time after the event at which the RHR system is put into service and CSD is achieved is 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months (10 days) for the LOOP scenario and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for the no-LOOP scenario. Primary-to-secondary leakage through the ASG is assumed to continue for the duration of the scenarios.

The schematic shown in Figure 4-2 provides a similar picture of the design basis MSLB involving a faulted MSSV. Hot full-power mass values were used for secondary side bulk liquid masses, consistent with the modeled steam releases. There are no releases to the Turbine Building.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 24 of 40 Figure 4-1: MSLB Turbine Building Release Model Reactor Coolant System 404,459 lbm Turbine Building 6E6 ft3 Affected Steam Generator (ASG) Bulk Liquid 109,160 lbm Credit is not taken for partitioning of iodine and moisture carryover of particulates.

Compartment mass reduced by 50% to aid the release of initial ASG inventory.

Initial Ci inventory is conserved by using source term fraction data from Table 4-9.

Intact Steam Generators (ISG)

Liquid - no tube uncovery 109,160 lbm/SG or 218,320 lbm Credit is taken for partitioning of iodine and moisture carryover of particulates.

ASG Primary-to-Secondary leak rate

- 2.896 lbm/min ISG Primary-to-Secondary leak rate - 5.445 lbm/min Exhaust to Environment ISG Bulk Liquid & RCS Activity Release through Safeties - See Table 4-14 Exhaust to Environment Release of Turbine Building Volume through Building Exhaust/Ventilation

- See Table 4-13 Exhaust to Control Room Direct Intake of Turbine Building Volume through Unfiltered Inleakage and Emergency Filtered Ventilation (available after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months through operator action) - See Figure 4-3 TDAFW to Environment ISG Bulk Liquid & RCS Activity Release through TDAFW Exhaust - See Table 4-15 Exhaust to Turbine Building ASG Bulk Liquid Initial Blowdown of Inventory in First 30 Minutes - See Table 4-13 RCS NG & progeny released directly to the ISG in an NG only case without reducing flow for partitioning or moisture carryover. NG case ISG exhaust flows to the environment used 100% efficient filters for P/E/O.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 25 of 40 Figure 4-2: MSLB Faulted MSSV Release Model Reactor Coolant System 404,459 lbm Affected Steam Generator (ASG) Bulk Liquid 109,160 lbm Credit is not taken for partitioning of iodine and moisture carryover of particulates.

Intact Steam Generators (ISG)

Liquid - no tube uncovery Credit is taken for partitioning of iodine and moisture carryover of particulates.

109,160 lbm/SG or 218,320 lbm ASG Primary-to-Secondary leak rate to Environment

- 2.896 lbm/min ISG Primary-to-Secondary leak rate - 5.445 lbm/min Exhaust to Environment ISG Bulk Liquid & RCS Activity Release through Safeties - See Table 4-17 Exhaust to Environment Rapid release of all curie content simulating ASG dry out early in the event. - See Table 4-16.

TDAFW to Environment ISG Bulk Liquid & RCS Activity Release through TDAFW Exhaust - See RCS NG & progeny released directly to the ISG in an NG only case without reducing flow for partitioning or moisture carryover. NG case ISG exhaust flows to the environment used 100% efficient filters for P/E/O.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 26 of 40 The Control Room modeled in this analysis contains three intake pathways: unfiltered inleakage (UFI), normal ventilation, and emergency ventilation. UFI draws from the environment for the duration of the event at 250 cfm. Normal ventilation operates while power is available, and the Control Room is not isolated at an unfiltered flow rate of 2800 cfm from the environment. Since the LOOP scenario assumes a coincident loss of offsite power, normal ventilation is not modeled in the LOOP cases. The no-LOOP scenario assumes the continued availability of offsite power. Therefore, normal ventilation operates until the Control Room is isolated after the start of the event as a result of a SI signal. The emergency ventilation system draws from the Turbine Building at 900 cfm after operator action places the system into service one hour after event initiation. Emergency ventilation remains in service for the remainder of the event.

The RADTRAD-NAI code calculates dose consequences based on the nuclide inventories transported between the compartments. The Control Room UFI and emergency intake would draw from the Turbine Building. The Turbine Building volume would be expected to include nuclides from the ASG blow down as well as from the ISG discharge to the environment. To model this scenario, the two sources were modeled as separate nuclide inventories: nuclides from the ASG blowdown in the Turbine Building and nuclides from the ISG discharge to the atmosphere. Thus, the RADTRAD-NAI model includes two UFI intake paths and two emergency intake paths that both use the nominal flow rates for the pathway. While the total flow rates are doubled for each path, the total nuclide inventory transported to the Control Room from each source is conserved. Before isolation with offsite power available (no-LOOP), the Control Room exhaust to the environment is modeled at 3050 cfm (2800 + 250). After isolation with no-LOOP, or for LOOP cases, the Control Room exhaust to the environment is modeled at 250 cfm before one hour and 1150 cfm (900 + 250) after the one hour. Figure 4-3 provides a visual representation of the Control Room model.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 27 of 40 Figure 4-3: MSLB Control Room Model Turbine Building 6E6 ft3 Control Room 2.30E5 ft3 Emergency Ventilation Filtered intake from Turbine Building 900 cfm, 1 - 720 Hours Emergency Ventilation Filtered intake from Environment 900 cfm, 1 - 720 Hours Unfiltered Inleakage Unfiltered intake from Environment 250 cfm, 0 - 720 Hours Unfiltered Inleakage Unfiltered intake from Turbine Building 250 cfm (after CR Isolation)

Normal Ventilation Unfiltered intake from Environment until CR Isolation (no-LOOP only) 2800 cfm, 0 second - CR isolation Control Room Exhaust Unfiltered release to Environment (equal to the sum of intake flows)

Environment

LBDCR No.25-017 Docket Nos. 50-338/339 Page 28 of 40 4.2.5.2 Release Rate Data Basic data and assumptions are shown in Table 4-1 and Table 4-2. Additional data include the mass and volumetric flow rates documented below.

Table 4-13: ASG Flow Rates to the Turbine Building and Turbine Building Volumetric Flow Rates to the Environment (Turbine Building Release w/LOOP)

Time (hrs)

Mass Flow Rate to Turbine Building, (lbm/min)

Turbine Building Volumetric Flow Rate to Environment (cfm) 0.0000E+00 3.3233E+05 5.0671E+06 1.5000E-03 2.9475E+05 4.4964E+06 3.1667E-03 2.4388E+05 3.7238E+06 3.9444E-03 1.7055E+05 2.6101E+06 4.8889E-03 1.1912E+05 1.8291E+06 6.7222E-03 8.8084E+04 1.3577E+06 8.8333E-03 6.9246E+04 1.0716E+06 1.1278E-02 5.5744E+04 8.6658E+05 1.4833E-02 4.1019E+04 6.4296E+05 2.2778E-02 3.0795E+04 4.8768E+05 3.3194E-02 2.3748E+04 3.8066E+05 7.5278E-02 9.1296E+03 1.5865E+05 5.0000E-01 0.0000E+00 2.0000E+04 3.0000E+00 0.0000E+00 2.0000E+04 2.4000E+02 0.0000E+00 2.0000E+04

LBDCR No.25-017 Docket Nos. 50-338/339 Page 29 of 40 Table 4-14: ISG Flow Rates to the Environment (Turbine Building Release w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 2.0278E-02 2.4704E+04 9.4722E-02 1.6776E+04 1.1236E-01 8.6739E+03 3.7111E-01 6.5316E+03 5.0000E-01 5.9990E+03 5.3000E-01 0.0000E+00 9.5694E-01 1.9202E+03 1.2028E+00 1.8838E+04 1.2069E+00 1.6856E+03 1.8097E+00 1.3598E+03 3.0000E+00 1.1700E+03 2.4000E+02 0.0000E+00 Note 1: These Steam Release Rates for all cases except those modeling Noble Gas releases were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the SG.

Table 4-15: TDAFW Exhaust Flow Rates (Turbine Building Release w/LOOP)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 4.3191E+02 8.1111E-03 4.8251E+02 1.1722E-02 5.1318E+02 1.5611E-02 5.4623E+02 7.0556E-02 5.2719E+02 5.0000E-01 5.2718E+02 3.0000E+00 5.2718E+02 2.4000E+02 0.0000E+00

LBDCR No.25-017 Docket Nos. 50-338/339 Page 30 of 40 Table 4-16: ASG Flow Rates to the Environment (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 1.2603E+04 4.7083E-02 8.5326E+03 9.5694E-02 6.6517E+03 2.5222E-01 6.3054E+03 5.0000E-01 7.5130E+03 7.3333E-01 4.6433E+03 7.5222E-01 1.6109E+03 7.7806E-01 3.2545E+01 3.0000E+00 2.7163E+01 2.4000E+02 0.0000E+00 Note 1: These Steam Release Rates were multiplied by an additional factor of 10 to simulate a rapid dry-out of the SG and a rapid release of all radionuclides.

Table 4-17: ISG Flow Rates to the Environment (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 3.0000E-03 4.1059E+04 1.1833E-02 1.1073E+04 3.5000E-02 2.1801E+03 1.1806E-01 2.7107E+04 1.2181E-01 1.6857E+01 1.3514E-01 2.7033E+04 1.3903E-01 5.6539E+03 2.1444E-01 2.5273E+04 2.1861E-01 1.6275E+01 2.3083E-01 2.6781E+04 2.3472E-01 0.0000E+00 2.4722E-01 2.6538E+04 2.5111E-01 0.0000E+00 2.6417E-01 2.6340E+04 2.6806E-01 4.3052E+03 3.1556E-01 2.6339E+04

LBDCR No.25-017 Docket Nos. 50-338/339 Page 31 of 40 Table 4-17: ISG Flow Rates to the Environment (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 3.1944E-01 0.0000E+00 3.3222E-01 2.6322E+04 3.3611E-01 3.4404E+01 3.4917E-01 2.6488E+04 3.5306E-01 7.7478E+00 3.6583E-01 2.6461E+04 3.6972E-01 7.8339E+01 3.8250E-01 2.6230E+04 3.8639E-01 0.0000E+00 3.9889E-01 2.6373E+04 4.0278E-01 3.5152E+01 4.1556E-01 2.6484E+04 4.1944E-01 0.0000E+00 4.3222E-01 2.6483E+04 4.3611E-01 5.4735E+03 5.0000E-01 1.1513E+04 5.3722E-01 6.3296E+03 5.6667E-01 4.5101E+03 7.8083E-01 0.0000E+00 8.3444E-01 2.3436E+04 8.3778E-01 1.8149E+03 1.3153E+00 1.6046E+03 1.8986E+00 1.3281E+03 3.0000E+00 1.1482E+03 2.4000E+02 0.0000E+00 Note 1: Steam Release Rates for all cases except those modeling Noble Gas releases were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the SG.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 32 of 40 Table 4-18: TDAFW Exhaust Flow Rates (Failed MSSV w/LOOP)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 4.8447E+02 1.5556E-03 5.3545E+02 1.7778E-02 5.2718E+02 5.0000E-01 5.2717E+02 3.0000E+00 5.2718E+02 2.4000E+02 0.0000E+00 4.2.6 Analysis Conservatisms The analysis includes several inputs and assumptions that are conservative and result in retained margin. Examples of these conservatisms are included below.

1. Event Timeline a) The assumed 10-day cooldown duration is beyond the anticipated time to cool down and depressurize the RCS.

b) This timeline was chosen to minimize time-critical operator actions and ensure Operations staff has ample time to recover equipment, choose the optimal long-term recovery strategy, and receive support from plant staff.

2. Dose Consequence Analysis Steam Release Rate (See Figure 4-4 for a visualization of expected conservatism) a) Steam Release input to the dose analyses represent 110% of the thermal hydraulic (T/H) calculation steam releases, and greater than 110% in the ASG to enhance the release of curies.

b) ISG Steam Release rates between 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 10 days are modeled as constant and equal to the ISG release rates corresponding to the core decay heat at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and terminated after 10 days (i.e., no credit for reduction in decay heat over time).

c) From day 7 to 10, use of the RHR system would result in heat removal from the RCS that would also lower the required amount of steaming from the ISGs. RHR heat removal was not credited to maximize steam releases and therefore the predicted dose consequences.

3. MCR Isolation timing extended beyond the predicted initiating signal and damper closure timing.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 33 of 40 Figure 4-4: Steam Release Conservatism Visualization 4.2.7 Analysis Results The total TEDE to the EAB, LPZ and Control Room from a MSLB is summarized in Table 4-19 for the concurrent and pre-accident spike. The concurrent spike results in the highest dose consequences for both offsite and the Control Room. The limiting cases for Control Room doses were Turbine Building releases w/LOOP for the pre-accident spike and MSSV w/LOOP for the concurrent spike. The limiting case for offsite doses was the faulted MSSV w/LOOP. The doses are within the acceptance criteria specified in RG 1.183 and 10 CFR 50.67.

Table 4-19: MSLB Accident Dose Summary Accident Location Dose Results (rem TEDE)

Acceptance Criteria (rem TEDE)

Concurrent Iodine Spike Control Room 2.9 5

EAB 0.5 2.5 LPZ 0.2 2.5 Pre-Accident Iodine Spike Control Room 1.5 5

EAB 0.1 25

LBDCR No.25-017 Docket Nos. 50-338/339 Page 34 of 40 Table 4-19: MSLB Accident Dose Summary Accident Location Dose Results (rem TEDE)

Acceptance Criteria (rem TEDE)

LPZ 0.1 25

5.0 REGULATORY EVALUATION

5.1 APPLICABLE REGULATORY REQUIREMENTS The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50.67, Accident source term applies to all holders of operating licenses issued prior to January 10, 1997, and holders of renewed licenses under part 54 of this chapter whose initial operating license was issued prior to January 10, 1997, who seek to revise the current accident source term used in their design basis radiological analyses. It requires a licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under 10 CFR 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

The NRC may issue the amendment under 10 CFR Part 50.67 only if the applicant's analysis demonstrates with reasonable assurance that:

i.

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

ii.

An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

iii.

Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 35 of 40 The regulations at 10 CFR 50, Appendix A, General Design Criteria for Nuclear Plants, Criterion 19, Control Room, states that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under 10 CFR 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in 10 CFR 50.2 for the duration of the accident.

The proposed change has no effect on the design of the control room or on operator radiation dose, as that protection is provided by other systems required by the Technical Specifications. The proposed change also has no effect on alternate control locations outside of the control room. Therefore, the only aspect of GDC 19 applicable to the proposed change is the criterion to design the control room from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. The proposed change has no effect on the design of the control room and the proposed actions will ensure that the control room temperature is maintained such that the plant may be operated safely from the control room.

Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents At Nuclear Power Reactors, Revision 0, provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. This guide establishes an

LBDCR No.25-017 Docket Nos. 50-338/339 Page 36 of 40 acceptable alternative source term (AST) and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

5.2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Virginia Electric and Power Company (Dominion Energy Virginia) proposes an update of the Main Steam Line Break (MSLB) Alternative Source Term (AST) dose consequence analysis for North Anna Power Station Units 1 and 2. The updated dose analysis was performed in response to industry operating experience and uses a timeline and steam release rates that bound the effect of a Reactor Coolant System (RCS) stagnant loop condition during cooldown to Cold Shutdown.

Dominion Energy Virginia has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment.

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

Response: No.

The proposed change to update the MSLB AST dose consequence analysis has been analyzed and has been determined to result in acceptable dose consequences, as it continues to meet the dose acceptance criteria specified in 10 CFR 50.67 and RG 1.183. The proposed change will not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition.

Updating the MSLB AST dose consequence analysis does not require any changes to any plant structures, systems, or components (SSCs), has no direct impact upon plant operation or configuration, and does not impact either the initiation of a currently evaluated accident or the mitigation of its consequences.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 37 of 40 The proposed change will not create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators not previously considered. Other than a change to the AST, there is no significant change to the other parameters within which the plant is normally operated, and no physical plant modifications are being made. Therefore, the possibility of a new or different type of accident is not created.

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

Response: No.

No design basis or safety limits are exceeded or altered by this change. Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed changes have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10CFR 50.67 and RG 1.183. The proposed changes will not result in in plant operation in a configuration outside the analyses or design basis and do not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, implementation of the proposed license amendment is safe and has no effect on plant operation. The proposed change makes no physical modifications to plant equipment or how equipment is operated or maintained. Consequently, Dominion Energy Virginia concludes the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Conclusion Based on the considerations presented above, there is reasonable assurance that: (1) the health and safety of the public will not be endangered by the demonstration that NAPS continues to meet applicable design criteria and safety analysis acceptance criteria, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 38 of 40

6.0 ENVIRONMENTAL CONSIDERATION

Dominion Energy Virginia has reviewed the proposed license amendment for environmental considerations in accordance with 10 CFR 51.22. The proposed license amendment 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 amendment meets the eligibility criterion for categorical exclusion from an environmental assessment as 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 amendment.

7.0 CONCLUSION

S The updated AST dose consequence analysis for the MSLB accident accounts for the effect of loop stagnation and continues to demonstrate the calculated personal dose results for the Control Room and at the EAB and LPZ meet regulatory requirements.

8.0 COMPUTER FILES An archive of computer files is associated with this LAR. The archive is comprised of three folders: 1) raw meteorological data, 2) ARCON96 input/output files, and 3)

RADTRAD input/output files. The ARCON96 files were used to generate atmospheric dispersion factors. The RADTRAD files were used to calculate dose consequences. The raw meteorological data files are included for general information.

The RADTRAD folder contains three subfolders: one subfolder of RADTRAD input/output files for each of the MSSV w/LOOP and Turbine Building release w/LOOP accident scenarios, and a subfolder of RADTRAD input files common to all the RADTRAD analyses. A README file contained within the archive provides additional details. Table 81 provides the quality assurance checksum for the data archive.

Table 81: Data Archive File Listing Data Archive Filename Checksum Data Files.zip 3389327054

LBDCR No.25-017 Docket Nos. 50-338/339 Page 39 of 40

9.0 REFERENCES

1. Regulatory Guide 1.183 Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, USNRC, Office of Nuclear Regulatory Research, July 2000.

2. 10 CFR 50.67, Accident Source Term.

3. Software - RADTRAD-NAI Version 1.3(QA), Numerical Applications Inc.

4. 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 19 - Control Room (GDC 19).

5. Letter SN 03-464, L. N. Hartz to USNRC, Proposed Technical Specification Changes, Implementation of Alternative Source Term, dated September 12, 2003 (ML032670821).

6. Letter from USNRC to D. A. Christian of Virginia Electric and Power Company, North Anna Power Station, Units 1 and 2 - Issuance of Amendments on Implementation of Alternative Source Term (TAC Nos. MC0776 and MC0777), dated June 15, 2005 (ML051590510).

7. Letter from USNRC to D. A. Christian of Virginia Electric and Power Company, Correction to Amendment Nos. 240 and 221, for North Anna Power Station (TAC Nos. MC0776 and MC0777), dated July 15, 2005 (ML051950547).

8. Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," EPA 520/1-88-020, Environment Protection Agency, 1988.

9. Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water and Soil," EPA 420-r-93-081, Environmental Protection Agency, 1993.

10. Callaway LER 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown, (ML18184A389), May 2018.

11. WANO SOER 99-1, Loss of Grid, December 1999, Addendum, December 2004.

12. NUREG-1784, Operating Experience Assessment - Effects of Grid Events on Nuclear Power Plant Performance, December 2003 (ML033530400).

13. IER L2-13-53, Loss of Off-Site Power (LOOP) Analysis, INPO Event Report.

14. INL/EXT-16-39575, Analysis of Loss-of-Offsite-Power Events 1987-2015, Idaho National Laboratory, July 2016.

15. ETE-SU-2017-0069, SPS Emergency Diesel Generator Mission Time, October 2017.

LBDCR No.25-017 Docket Nos. 50-338/339 Page 40 of 40

16. NUREG/CR-6890 Vol. 1, Reevaluation of Station Blackout Risk at Nuclear Power Plants, December 2005.

17. INL/RPT-22-68809, Analysis of Loss-of-Offsite-Power Events Update, Idaho National Laboratory, August 2022.

18. ETE-NAF-2023-0067, Dominion Energys NSA Position on Safe-Shutdown (HSD/CSD) Licensing Basis for the VA Units, February 2025.

19. NOTEBK-PRA-DOM-HR.3, Revision 4, "Dominion Probabilistic Risk Assessment Model Notebook Human Reliability Analysis Recovery of Loss of Offsite Power,"

November 2015.

20. Letter SN 03-464A, L. N. Hartz to USNRC, Proposed Technical Specification Changes, Implementation of Alternative Source Term, Request for Additional Information, dated November 20, 2003 (ML033350516).

21. Letter SN 23-045, J. E. Holloway to USNRC, Proposed Emergency Plan Revision -

Relocation of the Technical Support Center (TSC), Supplemental Information - LOCA Dose Calculation Summary Report and Elimination of CO2 Fire Suppression System, dated June 27, 2023 (ML23192A215).

22. NUREG/CR-6331, Rev. 1, Atmospheric Relative Concentrations in Building Wakes, ARCON96, USNRC, 1997.

23. Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments in Nuclear Power Plants, June 2003.

24. TTG-101-2022, Rev. 0, Multi-Tracer Testing at Dominions North Anna Station for Air In-Leakage Determination, Brookhaven National Laboratory, May 2022.

Serial No.25-017 Docket Nos. 50-280/281 LICENSE AMENDMENT REQUEST UPDATED MAIN STEAM LINE BREAK ALTERNATE SOURCE TERM DOSE CONSEQUENCE ANALYSIS Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Serial No.25-017 Docket Nos. 50-280/281 Page 1 of 41 LICENSE AMENDMENT REQUEST UPDATED MAIN STEAM LINE BREAK ALTERNATE SOURCE TERM DOSE CONSEQUENCE ANALYSIS SURRY POWER STATION UNITS 1 AND 2

1.0 INTRODUCTION AND BACKGROUND

In 2018, the Callaway Nuclear Plant issued License Event Report (LER) 2018-002-00

[Reference 9], which identified that following a Main Steam Line Break (MSLB) with a loss of offsite power (LOOP), an asymmetric natural circulation cooldown (ANCC) could result in Reactor Coolant System (RCS) flow stagnation in the affected loop. Flow stagnation would then have the effect of extending the timeline to cooldown and depressurize the RCS following a MSLB.

Dominion Energy evaluated the effect of extending the timeline to cooldown and depressurize the RCS following a MSLB and determined that the dose consequence analyses for the MSLB Event at Surry Power Station (SPS) was affected. As a result, a new bounding timeline was determined, and new dose consequence analyses were performed. The following sections describe the stagnant loop phenomenon, as well as Dominion Energys bases for maintaining safe shutdown (i.e., Hot Shutdown definition in the Technical Specifications (TS)), 10-day steam release duration, and radiological release termination.

The proposed LAR scope is associated with only the MSLB Dose Consequence analysis and does not include the effects (if any) of a stagnant loop condition on other Design Basis Accidents.

For other scenarios (e.g., Steam Generator Tube Rupture), Dominion Energy continues to investigate the effects and will submit separate LARs if required.

Stagnant Loop Phenomenon During a MSLB event, the primary mitigating actions are to isolate feedwater (FW) and Auxiliary Feedwater (AFW), as well as isolate the steaming flow paths for the faulted steam generator (SG). By isolating the SG, the affected RCS loop becomes inactive. Once the SG is isolated and the RCS is stabilized, the unit has reached a safe shutdown condition. After the establishment of safe shutdown, the following phase is to perform an RCS cooldown and depressurization for long-term recovery.

When natural circulation is established with at least one inactive loop, it is known as asymmetric natural circulation (ANC). When ANC occurs, flow through the inactive loop may cease, resulting in a stagnant loop condition. When performing an ANC cooldown with a stagnant loop, the operators control the cooldown of the bulk RCS fluid, but the stagnant loop remains hot since heat transfer in that loop has ceased. Unless flow is re-established permanently or temporarily, the affected RCS loop will remain at elevated temperatures for an extended period of time. This becomes an issue when RCS depressurization to Residual Heat Removal (RHR) system entry

Serial No.25-017 Docket Nos. 50-280/281 Page 2 of 41 conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization.

In this condition, the time required to complete a cooldown to Cold Shutdown (CSD) conditions could be sufficiently extended beyond the current safety analyses assumptions (i.e., 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months for SPS). When evaluating the effects of the stagnant loop condition, as well as potential operational strategies to expedite the cooldown, Dominion Energy determined an extended qualitative bounding timeline was more appropriate. The bounding timeline allows Operations to establish a safe shutdown condition and then perform either an expedited ANCC or hold at safe shutdown conditions for up to 7 days. Using this assumption, the unit may be held at safe shutdown until either offsite power is restored (allowing start of a Reactor Coolant Pump (RCP) which eliminates the stagnant loop concern), the faulted SG is recovered (allowing a symmetric plant cooldown which eliminates the stagnant loop concern), or a planned ANCC is performed.

The timeline developed separates expedited ANCC strategies from the dose consequence analyses (i.e., the expedited strategies are not credited in the analysis). By analyzing such a large timeline for steam release, the Operations team and Emergency Response Organization (ERO) will have ample time to recover equipment and reduce unnecessary expedited strategies.

This approach of remaining at safe shutdown (i.e., Hot Shutdown (HSD)) conditions for a 10-day period (i.e., 7-day period represents LOOP duration time with a 3-day forced cooldown to RHR entry conditions and RHR cooldown CSD) was used to analyze the dose consequences of an extended MSLB event. It is important to note that the analyses do not credit or require restoration of offsite power as the primary success path for the safety analyses, instead it is chosen as a bounding assumption which envelopes all scenarios that result in a cooldown in less than 10 days.

Basis for maintaining Hot Shutdown mode of operation:

SPS is designed and licensed to establish and maintain a safe shutdown condition following a design-basis accident. More explicitly, SPS is designed and licensed to maintain HSD mode of operation as a safe shutdown mode using the SGs and atmospheric dump valves (SG Power Operated Relief Valves (PORVs)) to remove decay and sensible heat if the condenser steam dump valves are not available. (Note the SG PORVs steam release function is not credited in the safety analyses for overpressure mitigation; therefore, the intact SGs (ISGs) will limit and control pressure utilizing the Main Steam Safety Valves (MSSVs)) [Reference 17]. A subsequent cooldown can be performed using both the ISGs and non-credited or non-safety grade equipment (i.e., steam generator PORVs and RHR system) when available to complete the cooldown to less than 200°F. It is important to note that since SPS is a Hot Shutdown plant by design and licensing basis, it does not have single-failure proof safety related equipment installed to reach CSD. Therefore, to comply with Regulatory Guide 1.183, Revision 0

[Reference 1], the dose consequence analysis must model steam releases until the unit reaches cold shutdown, and credit must be taken for the use of non-safety related equipment. It is also known that sufficient, alternate water supplies may be needed to support AFW system operation if the plant is required to stay in a safe shutdown condition for an extended period.

Serial No.25-017 Docket Nos. 50-280/281 Page 3 of 41 In summary, the updated analyses model SPS maintaining safe shutdown in the HSD condition for an extended period, with a subsequent cooldown to CSD conditions when appropriate equipment to perform the cooldown becomes available and as directed by senior station personnel. The subsequent cooldown may credit non-safety grade structures, systems, and components (SSCs) to achieve the CSD mode of operation.

Basis for 10-day radiological release duration:

A review of industry operating experience (OE) shows that longer full LOOP recovery durations are typically caused by widespread damage due to severe weather events such as tornadoes, hurricanes, and earthquakes. Nuclear plants were designed to withstand natural events based on the historical and environmental data available at the time. However, the industry now has more than 40-50 years of operating history that should also be considered when evaluating LOOP duration times.

The OE shows that for the vast majority of LOOP events, offsite power was restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from initiation. Comprehensive LOOP studies performed by Idaho National Laboratory (INL) and Institute of Nuclear Power Operations (INPO) draw similar conclusions [References 10-13]. The studies concluded the vast majority of events are caused by equipment and human performance related issues and are recovered in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . More extensive damage and longer recovery times were more often due to large scale severe weather events. The studies also concluded that LOOP recovery time trended upward due to additional time being taken for investigation prior to restoring the preferred power source [Reference 14].

The longest recorded LOOP recovery time was at Turkey Point in 1992, which took approximately 6 1/2 days to recover offsite power following Category 5 Hurricane Andrew

[References 15 and 16]. In 2017, Category 4 Hurricane Irma, struck southern Florida; however, grid power to Turkey Point was never lost. This was due in part to improvements in alternating current (AC) power reliability made on and offsite. Likewise, Dominion Energys Virginia Nuclear Fleet has improved their switchyard design based on this and other relevant OE in recent years, thereby increasing the AC reliability from when the plant was first built and licensed. Probability Risk Assessment (PRA) analysis for SPS determined the mean recovery time for all LOOP events that occurred during power operation is 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and 21 minutes, while the mean recovery time specifically for weather-related LOOP events is 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months and 11 minutes [Reference 18].

Based on the longest LOOP recovery time being 6 1/2 days during the last 39 years of nuclear industry data, a 7-day LOOP recovery time is considered reasonable.

Note the licensing bases for SPS does not require the MSLB event be considered coincident with a severe weather event. Therefore, the timeline for expected recovery of offsite power does not need to consider severe weather events. For the purposes of conservatism, the timeline used by Dominion Energy was extended to exceed even those of severe weather-related LOOP durations.

Based on industry weather related LOOP event recovery times, SPS redundancy of offsite sources, General Design Criteria (GDC) 17 independence of electrical design, and Dominion Energy's demonstrated emergency preparedness and restoration capabilities, a 7-day LOOP

Serial No.25-017 Docket Nos. 50-280/281 Page 4 of 41 recovery time for determining radiological consequences in a safe shutdown condition is reasonable. Therefore, in absence of a regulatory or industry definition, a reasonable LOOP recovery time for determining radiological consequences while at a safe shutdown condition was selected for SPS. At this point, the recovery of offsite power or recovery of the faulted Steam Generator will occur to allow the operators to perform a forced or symmetric cooldown to CSD conditions which eliminates stagnant loop and ANCC concerns.

Once an offsite feeder is restored, an RCP can be started allowing a forced flow cooldown and/or the faulted SG is recovered to allow a symmetric cooldown to RHR entry conditions (assumed 1 day duration), followed by an RHR cooldown to CSD assumed 1 day duration), and cooldown of the ISGs to < 212 °F (assumed 1 day duration). Therefore, the dose consequences for a MSLB with a LOOP has the potential to last for 10 days, avoiding the stagnant loop and ANCC complexity. Its important to note that specific timing of each phase is irrelevant to the analysis; the total duration of 10 days is the dominating contributor.

Basis for radiological release termination:

Dominion Energy adopted and incorporated the AST methodology for dose consequences at SPS in 2002 [Reference 19]. Reference 1, Table 6, states that the analysis release duration for a PWR MSLB is until Cold Shutdown is established (i.e., RCS average coolant temperature 200°F). Appendix E of Reference 1 states the following under Transport 5.3:

The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Dominion Energys position for the MSLB dose analysis is to assume the steam release rate that was calculated at four hours is extended to 10 days, at which point the release is considered terminated. Using this conservative 10-day approach on an MSLB event to CSD provides a bounding dose consequence, which allows the operators to select the desired path based on actual plant parameters, equipment availability, and station leadership direction to achieve CSD (termination of the radiological consequence). Increasing the extended cooldown timeline to avoid an ANCC situation would result in increased dose consequences that exceed the more than minimal threshold for 10 CFR 50.59.

Figure 1-1 demonstrates a simplified graphical representation of the three distinct phases considered for the MSLB dose analysis timeline. This timeline is based on the discussions above for each phase.

Serial No.25-017 Docket Nos. 50-280/281 Page 5 of 41 Figure 1-1: Phases of MSLB Accident Mitigation 2.0 RADIOLOGICAL CONSEQUENCES INTRODUCTION This report describes the evaluations conducted to assess offsite doses and Control Room habitability at Surry Power Station (SPS) following a postulated MSLB design basis accident per Regulatory Guide (RG) 1.183, Revision 0 [Reference 1]. The accident source term discussed in Reference 1 is herein referred to as the Alternative Source Term (AST). The evaluations have employed the detailed methodology contained in RG 1.183 for use in design basis accident analyses for the AST. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 [Reference 2] or the supplemental guidance in RG 1.183.

This application includes the following key elements:

Increase the time required for cooldown; Adding an MSSV failure scenario.

3.0 CURRENT LICENSING BASIS

SUMMARY

The current MSLB design basis radiological analysis which appears in SPS Updated Final Safety Analysis Report (UFSAR) Section 14.3.2 was submitted for approval in Reference [5] and approved in Reference [6]. The analysis was performed using the RADTRAD-NAI code based on a core inventory derived with the ORIGEN-ARP code.

Event Mitigation

Use of only Safety Related (SR)

Equipment that is protected by Technical Specifications

MSLB Isolation

RCS Stablization at Safe Shutdown Conditions Hold at Safe Shutdown

Long-Term Recovery

Follows Transient Mitigation

May result in makeup to SR water sources from non-safety (NS) sources

May be up to 7 days Cooldown /

Depressurization

Long-Term Recovery

Cooldown to CSD IAW Dose Consequence RG Requirements

Uses NS Equipment

May take up to 3 additional days

Serial No.25-017 Docket Nos. 50-280/281 Page 6 of 41 4.0 ANALYSIS ASSUMPTIONS AND KEY PARAMETER VALUES This section describes the general analysis approach and presents analysis parameters, Table 4-1, and assumptions, Table 4-2. The analysis parameters and assumptions presented in Table 4-1 and Table 4-2 have not changed from the values documented in References [5] and [6].

The dose analyses documented in this application employ the Total Effective Dose Equivalent (TEDE) calculation method, as specified in Reference [1] for AST applications. The TEDE is determined at the Exclusion Area Boundary (EAB) for the worst 2-hour interval. TEDE values for individuals at the Low Population Zone (LPZ) and for SPS Control Room personnel are calculated for the assumed duration of the event.

The TEDE concept is defined to be the Deep Dose Equivalent, DDE (from external exposure) plus the Committed Effective Dose Equivalent, CEDE (from internal exposure). In this manner, TEDE assesses the impact of relevant nuclides upon body organs, in contrast with the previous single, critical organ (thyroid) concept for assessing internal exposure. CEDE dose conversion factors were taken from Table 2.1 of Federal Guidance Report (FGR) 11, Reference [7], per Section 4.1.2 of Reference [1]. The DDE is nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE is used instead of DDE in determining the contribution of external dose to the TEDE. EDE dose conversion factors were taken from Table III.1 of FGR 12, Reference [8], per Section 4.1.4 of Reference [1].

Table 4-1: Analysis Parameters Parameter Value Control Room Emergency Intake Filtered Flow 900 cfm Control Room Normal Intake Unfiltered Flow 3300 cfm Control Room Volume 223,000 ft3 Control Room Unfiltered Inleakage 250 cfm Main Control Room Emergency Fan Actuation Time 60 minutes after the SI (modeled as 60 minutes after the start of a MSLB) 50% of Turbine Building Free Air Volume 3E6 ft3

Serial No.25-017 Docket Nos. 50-280/281 Page 7 of 41 Table 4-1: Analysis Parameters Parameter Value RCS total leak rate to 3 Steam Generators 1 gallon/min RCS Leakage Liquid Density 62.4 lbm/ft3 Control Room Filter Efficiencies Product Efficiency (%)

Elemental Iodine 90 Methyl Iodine (organic) 70 Aerosol (particulate) 99 Moisture Carryover in Intact SG 1%

Iodine Partitioning Source Fractional Value Affected Steam Generator Break Flow 1.0 ISG 0.01 Normal CR Intake Atmospheric Dispersion Factor (MSSV)

Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3) 0-2 9.88E-03 1.98E-03 2-8 7.25E-03 1.45E-03 8-24 3.21E-03 6.42E-04 24-96 2.04E-03 4.08E-04 96-720 1.45E-03 2.90E-04 Emergency CR Intake Atmospheric Dispersion Factor (MSSV)

Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3) 0-2 2.86E-03 5.72E-04 2-8 2.29E-03 4.58E-04 8-24 8.90E-04 1.78E-04 24-96 6.49E-04 1.30E-04 96-720 4.88E-04 9.76E-05

Serial No.25-017 Docket Nos. 50-280/281 Page 8 of 41 Table 4-1: Analysis Parameters Parameter Value Normal CR Intake Atmospheric Dispersion Factor (Turbine Driven Auxiliary Feedwater (TDAFW)

Turbine Exhaust)

Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3) 0-2 1.43E-02 2.86E-03 2-8 1.11E-02 2.22E-03 8-24 4.46E-03 8.92E-04 24-96 3.29E-03 6.58E-04 96-720 2.40E-03 4.80E-04 Emergency CR Intake Atmospheric Dispersion Factor (TDAFW Turbine Exhaust)

Time (hr)

Dispersion Factor (s/m3)

Dispersion Factor ÷ 5 (s/m3) 0-2 5.00E-03 1.00E-03 2-8 3.80E-03 7.60E-04 8-24 1.52E-03 3.04E-04 24-96 1.10E-03 2.20E-04 96-720 8.30E-04 1.66E-04 Minimum Steam Velocity Required for Application of Atmospheric Dispersion Factor ÷ 5 5 x 95th Percentile Wind Speed (mph)

Release Point 59.3 PORV/MSSV exhaust 42.7 TDAFW exhaust EAB Atmospheric Dispersion Factor Time (hrs)

Dispersion Factor (s/m3) 0 - 720 1.19E-03 LPZ Atmospheric Dispersion Factor Time (hrs)

Dispersion Factor (s/m3) 0 - 2 5.73E-05 2 - 8 5.73E-05 8 - 24 3.89E-05 24 - 96 1.68E-05 96 - 720 5.05E-06 Control Room Breathing Rate Time (hrs)

Breathing Rate (m3/s) 0 - 720 3.5E-04 EAB Breathing Rate Time (hrs)

Breathing Rate (m3/s) 0 - 720 3.5E-04

Serial No.25-017 Docket Nos. 50-280/281 Page 9 of 41 Table 4-1: Analysis Parameters Parameter Value LPZ Breathing Rate Time (hrs)

Breathing Rate (m3/s) 0 - 8 3.5E-04 8 - 24 1.8E-04 24 - 720 2.3E-04 Control Room Occupancy Factors Time (hrs)

Occupancy Factor 0 - 8 1.0 8 - 24 1.0 24 - 96 0.6 96 - 720 0.4 Iodine Chemical Form Product Fractional Distribution Elemental Iodine 0.97 Methyl Iodine (organic) 0.03 Aerosol (particulate) 0.0 RCS TS Limit for Dose Equivalent I-131 1.0 µCi/gm RCS TS Limit for Dose Equivalent Xe-133 234 µCi/gm Steam Generator TS Limit for Dose Equivalent I-131 0.1 µCi/gm

Serial No.25-017 Docket Nos. 50-280/281 Page 10 of 41 Table 4-2: Analysis Assumptions Assumption Description Turbine Building Exhaust Rate 0.2 volumes/hr (Note 1)

Note 1: This assumption does not affect the analyses associated with the faulted MSSV. The case modeling a break in the Turbine Building uses the low turnover rate to maximize the Control Room dose.

4.1 Proposed Licensing Basis Changes This section provides a summary description of the key proposed licensing basis changes that are justified with the revised SPS AST MSLB analysis. This LAR is being submitted for prior NRC review and approval pursuant to the requirements of 10 CFR 50.59 which specifies that a departure from a method described in the UFSAR, such as the design basis radiological consequence analyses, shall be submitted for approval unless the changes to the elements of the method meet certain requirements.

The proposed changes for a MSLB radiological event are:

1) An increase of the cooldown times associated with LOOP conditions,

2) Adding a malfunctioning MSSV as a release path (due to extended release time and potential for new limiting scenario).

The proposed changes have been analyzed and result in acceptable consequences, thereby meeting the criteria as specified in References [1] and [2]; however, they did not meet the requirements for implementation under 10 CFR 50.59 without prior NRC approval.

Serial No.25-017 Docket Nos. 50-280/281 Page 11 of 41 Table 4-3: Comparison of Proposed Changes Parameter CLB Value Proposed Value Reason for Change Control Room Isolation, hours (seconds)

LOOP 0.01083 (39)

No-LOOP 0.01000 (36)

Turbine Building Releases 0.0167 (60)

Faulted MSSV 0.072 (260)

Proposed Control Room isolation times include Safety Injection (SI) signal, Control Room Damper closure time, and retained margin.

CLB isolation times were conservative but included no margin.

SG Liquid Mass per SG 93,261 lbm /SG (Intact) 154,490 lbm /SG (Faulted) 105,700 lbm /SG Hot Full Power (HFP) conditions, consistent with the source document for steam flows RCS Liquid Mass 406,300 lbm 401,790 lbm HFP conditions, consistent with the source document for steam flows Release Duration, hours LOOP 38 No-LOOP 12.5 LOOP 240 No-LOOP 12.5 Recovery from Stagnant Loop after 10 days

Serial No.25-017 Docket Nos. 50-280/281 Page 12 of 41 4.2 Main Steam Line Break (MSLB) Reanalysis This application involves the reanalysis of the design basis radiological analyses for the MSLB. The calculated radiological consequences are compared with the acceptance criteria provided in 10 CFR 50.67(b)(2), as clarified per the additional guidance in RG 1.183 for events with a higher probability of occurrence.

Dose calculations are performed at the EAB for the worst 2-hour period, and for the LPZ and SPS Control Room for the duration of the accident. The radiological dose consequence calculations were performed with the RADTRAD-NAI computer code system [Reference 3]. The applicable dose acceptance criteria are provided in Table 4-

4.

Table 4-4: MSLB Accident Dose Acceptance Criteria Accident Control Room (Note 1)

EAB & LPZ Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Concurrent Iodine Spike 5 rem TEDE 2.5 rem TEDE Note 1: Based on 10 CFR 50.67 and 10 CFR 50, Appendix A, GDC 19 [Reference 4].

This analysis includes doses associated with the releases of radioactive material initially present in primary and secondary liquids at maximum allowable TS concentrations plus iodine spiking scenarios. No fuel failure is expected. Doses were calculated at the EAB, LPZ, and Control Room. The methodology used to evaluate the dose consequences resulting from the MSLB is consistent with RG 1.183 [Reference 1].

4.2.1 MSLB Scenario Description The MSLB accident begins with a break in one of the Main Steam lines leading from an affected SG (ASG) to the turbine. To maximize doses, break scenarios are assessed as follows: 1) a break in the Turbine Building, and 2) a faulted MSSV.

4.2.1.1 Turbine Building Release The ASG blows down into the Turbine Building for 30 minutes, after which it is isolated (via an existing time-critical operator action) and releases from this pathway stop. Two scenarios are modeled for the MSLB: Loss of Offsite Power (LOOP) and Offsite Power Available (no-LOOP). Each scenario uses steam release rates specific to the scenario.

In addition, both scenarios model Turbine Building exhaust differently. The LOOP scenario models the Turbine Building exhaust fans as not having the power required to operate. Therefore, natural circulation (0.2 building-volumes per hour) is modeled for the LOOP scenario. The no-LOOP scenario models the Turbine Building exhaust flow rate at

Serial No.25-017 Docket Nos. 50-280/281 Page 13 of 41 the maximum capability of the Turbine Building exhaust fans (12 building-volumes per hour). The total modeled flow rate from the Turbine Building includes the ASG blow down flow rates. To maximize dose consequences, neither scenario credits condenser availability.

Primary and secondary activity discharged from the ASG is released directly to the Turbine Building without mitigation. Primary-to-secondary leakage in the ASG is assumed at 500 gpd (2.90 lbm/min); total primary-to-secondary leakage modeled from the SGs is 1440 gpd (1 gpm accident induced leak rate specified in TS 6.4.Q.2.b). The TS RCS Operational Leakage primary-to-secondary leak rate is 150 gpd through any one SG as specified in TS 3.1.C.1.d.

The primary system is cooled down through the release of steam from the two ISGs.

Steam releases from the ISGs are modeled to occur through the MSSVs and TDAFW turbine exhaust. Modeling a release through the MSSVs bounds the use of either the PORVs or condenser steam dumps. The ISGs are modeled to have a total leakage rate equal to 940 gpd, which is the remaining primary-to-secondary leakage (1440 gpd - 500 gpd). ISG steaming will continue until sufficient cooldown allows the RHR system to be placed into service. The times for the RHR system to be placed into service are modeled as 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months and 12.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the initiation of the event for the LOOP and no-LOOP scenarios, respectively.

4.2.1.2 Faulted MSSV Release ASG vents directly to the atmosphere for the duration of the event. The scenario is modeled with a LOOP. The no-LOOP scenario is non-limiting because SI would not occur; further, with offsite power available a normal cooldown and depressurization would occur.

There are no releases to the Turbine Building. To maximize dose consequences, the condenser is not credited as being available.

AFW additions to the ASG may occur for up to 30 minutes. The limiting case was an immediate termination of AFW water followed quickly by steam generator dry-out. Steam flow to the environment was increased by a factor of 10 to ensure all activity in the ASG was quickly released to the environment.

Primary and secondary activity discharged from the ASG is released directly to the environment, modeling steam generator dry-out. Primary-to-secondary leakage in the ASG is assumed at 500 gpd; total primary-to-secondary leakage modeled from the SGs is 1440 gpd (1 gpm accident induced leak rate specified in TS 6.4.Q.2.b). The TS RCS Operational Leakage primary-to-secondary leak rate is 150 gpd through any one SG as specified in TS 3.1.C.1.d.

Cooldown of the primary system is through the release of steam from the two ISGs. Steam releases from the ISGs are modeled to occur through the MSSVs and TDAFW turbine exhaust. An actual plant cooldown would require steam release through the SG PORVs or condenser steam dumps; however, the modeling assumption (in the dose

Serial No.25-017 Docket Nos. 50-280/281 Page 14 of 41 consequence analysis) of MSSV is representative for the PORVs and conservative relative to condenser steam dumps. The ISGs are modeled to have a total leakage rate equal to 940 gpd, which is the remaining primary-to-secondary leakage (1440 gpd - 500 gpd). ISG steaming will continue until sufficient cooldown allows the RHR system to be placed into service. The time for the RHR system to be placed into service is modeled as 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months for the LOOP scenario.

4.2.2 MSLB Source Term Definition The primary coolant source term definition is unchanged from the CLB (Reference 5).

The analysis of the MSLB accident indicates that no fuel rod failures occur as a result of the transient. Thus, radioactive material releases during the event are determined by assuming the radionuclide concentrations initially present in primary and secondary liquid are at maximum TS limits plus iodine spiking. In accordance with RG 1.183, Appendix E, two independent cases are evaluated. Case one assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike.

The MSLB analysis uses the primary coolant concentration shown in Table 4-5 and the pre-accident iodine spike source term shown in Table 4-6. Initial secondary side liquid concentration is 10% of the primary concentration (based on the ratio of RCS and SG TS Dose Equivalent (DE) I-131 limits) without noble gases. Initial secondary side steam concentrations are negligible. The MSLB analysis models the concurrent iodine spike, as shown in Table 4-7, which corresponds to an accident-initiated value 500 times the equilibrium appearance rate persisting for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Kr-83m 1.60E-01 Kr-85m 5.79E-01 Kr-85 2.09E+00 Kr-87 3.86E-01 Kr-88 1.08E+00 Xe-131m 1.23E+00 Xe-133m 1.51E+00 Xe-133 1.02E+02

Serial No.25-017 Docket Nos. 50-280/281 Page 15 of 41 Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Xe-135m 3.90E-01 Xe-135 4.09E+00 Xe-138 2.75E-01 Br-83 2.86E-02 Br-84 1.54E-02 I-129 2.75E-08 I-130 1.16E-02 I-131 7.42E-01 I-132 3.85E-01 I-133 1.25E+00 I-134 2.50E-01 I-135 8.23E-01 Se-81 2.52E-07 Se-83 3.46E-07 Rb-86 1.03E-02 Rb-88 1.12E+00 Rb-89 6.72E-02 Sr-89 9.17E-04 Sr-90 5.67E-05 Sr-91 4.70E-04 Sr-92 3.82E-04

Serial No.25-017 Docket Nos. 50-280/281 Page 16 of 41 Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Y-90 7.01E-05 Y-91m 2.79E-04 Y-91 1.03E-03 Y-92 4.30E-04 Y-93 2.30E-04 Y-94 1.19E-05 Y-95 4.93E-06 Zr-95 1.62E-04 Zr-97 1.17E-04 Nb-95m 1.82E-06 Nb-95 1.64E-04 Nb-97m 1.10E-04 Nb-97 1.23E-04 Mo-99 1.09E+00 Mo-101 8.75E-03 Tc-99m 4.69E-01 Tc-101 8.39E-03 Ru-103 1.48E-04 Ru-105 4.63E-05 Ru-106 5.11E-05 Rh-103m 1.49E-04

Serial No.25-017 Docket Nos. 50-280/281 Page 17 of 41 Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Rh-105 9.76E-05 Rh-106 5.70E-05 Rh-107 4.42E-06 Sn-127 9.15E-07 Sn-128 2.00E-06 Sn-130 3.48E-07 Sb-127 7.79E-06 Sb-129 1.33E-05 Sb-131 4.78E-06 Te-125m 1.03E-04 Te-127m 8.32E-04 Te-127 3.47E-03 Te-129m 3.62E-03 Te-129 4.59E-03 Te-131m 1.00E-02 Te-131 4.68E-03 Te-132 7.96E-02 Te-133m 7.76E-03 Te-133 3.57E-03 Te-134 1.20E-02 Cs-134m 1.47E-02

Serial No.25-017 Docket Nos. 50-280/281 Page 18 of 41 Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Cs-134 1.11E+00 Cs-136 3.05E-01 Cs-137 8.24E-01 Cs-138 4.21E-01 Ba-137m 7.72E-01 Ba-139 3.14E-02 Ba-140 1.14E-03 Ba-141 5.07E-05 Ba-142 7.85E-05 La-140 3.16E-04 La-141 9.51E-05 La-142 9.65E-05 La-143 5.92E-06 Ce-141 1.58E-04 Ce-143 1.29E-04 Ce-144 1.22E-04 Pr-143 1.49E-04 Pr-144 1.23E-04 Pr-145 5.27E-05 Nd-147 6.24E-05 Nd-149 8.62E-06

Serial No.25-017 Docket Nos. 50-280/281 Page 19 of 41 Table 4-5: Primary (RCS) Coolant Concentration Nuclide Primary Concentration (Ci/gm)

Nd-151 6.72E-07 Pm-147 2.96E-05 Pm-149 5.21E-05 Pm-151 1.61E-05 Sm-151 1.62E-07 Na-24 1.41E-01 Cr-51 9.30E-03 Mn-54 4.80E-03 Fe-55 3.60E-03 Fe-59 9.00E-04 Co-58 1.38E-02 Co-60 1.59E-03 Zn-65 1.53E-03 Np-239 6.60E-03 H-3 2.50E+00

Serial No.25-017 Docket Nos. 50-280/281 Page 20 of 41 Table 4-6: Pre-accident Iodine Spike - 10 µCi/gm DE I-131 (see Note 1)

Nuclide Concentration

(µCi/gm)

I-131 7.42E+00 I-132 3.85E+00 I-133 1.25E+01 I-134 2.50E+00 I-135 8.23E+00 Note 1: The spike was implemented as shown in Table 4-8. The spike was modeled using a TS iodine source with a release fraction of 1 and an iodine spike source with an iodine release fraction of 9. In this manner the complete pre-accident iodine spike of 10 µCi/gm DE I-131 was modeled.

Table 4-7: Concurrent Iodine Spike MSLB RCS Concentration Nuclide Spike = 500 x Appearance Rate (Ci/hr)

Iodine Released 8-hour Spike Duration (Ci)

I-131 1.12E+04 8.96E+04 I-132 1.55E+04 1.24E+05 I-133 2.20E+04 1.76E+05 I-134 2.03E+04 1.62E+05 I-135 1.94E+04 1.55E+05 The RCS concentrations in Table 4-5 were used to build a nuclide inventory file (NIF) that was used to model the RCS, SG liquid, and pre-accident spike activities. To obtain total radioisotope inventories for each compartment from the RCS 1.0 µCi/gm DE I-131 NIF, the concentrations from the NIF were converted using a power level (within RADTRAD-NAI) derived from a combination of compartment masses, unit conversion factors, and

Serial No.25-017 Docket Nos. 50-280/281 Page 21 of 41 other parameters that affect radioisotope inventories in a compartment. For example, the source term fraction entry for the RCS compartment is determined by multiplying the mass of the RCS (lbm) by conversion factors for gm/lbm and Ci/µCi:

RCS = MRCS (lbm) x 453.59237 (gm/lbm) x 1.00E-06 (Ci/µCi)

Iodine and particulate source term fractions of the initial SG liquid compartments for the ASG and ISG were determined using the mass of the SG liquid in the same manner.

However, the SG sources also used a release fraction of 0.1 to reflect the differences in TSs. Additionally, the source term is split between the affected and ISGs based on their individual liquid masses.

Table 4-8 contains the parameters necessary to model the source term fractions. The use of the data developed in Table 4-8 in conjunction with the RCS activity NIF, based on Table 4-5, allows the RCS, SG liquid, and the two spike activity releases to be modeled in multiple RADTRAD-NAI cases.

Table 4-8: Source Term Fraction Values by Compartment and Source Compartment / Source Compartment Source Term Fraction RFT Release Fraction Values PSF Power Level (Scaling Factor)

RCS TS Source 1

1 182.25 RCS Noble Gases Only 1

1 182.25 RCS Pre-Incident Iodine Spike 1

9 for Iodine, 0 for others 182.25 RCS Concurrent Iodine Spike 1

1 for Iodine, 0 for others 1

Steam Generator TS Source 0.667 for ISG 0.333 for ASG 0.1 0 for Noble Gases 143.83

Serial No.25-017 Docket Nos. 50-280/281 Page 22 of 41 4.2.3 MSLB Release Transport 4.2.3.1 Turbine Building Release For the ASG, the release pathway is assumed to be directly into the Turbine Building with no credit taken for partitioning or scrubbing of the SG liquid. From the Turbine Building, the activity is assumed to pass into the Control Room via emergency (filtered) intakes and unfiltered inleakage as well as pass into the environment through vents, louvers, and other openings located around the Turbine Building. The ASG will release activity into the Turbine Building until isolated at 30 minutes. The release from the Turbine Building is determined by each scenario. In the LOOP scenario, the Turbine Building ventilation system is not operating due to offsite power being unavailable. In the no-LOOP scenario, the system is determined to be energized and is modeled as operating at its maximum capacity. The models use 0.2 and 12 building-volumes per hour for each scenario, respectively.

The ASG transport model that is utilized for noble gases, iodine, and particulates is consistent with Appendix E of RG 1.183. During the first 30 minutes of the event while the ASG is blowing down, radioactivity in the bulk liquid is released without reduction for partitioning or scrubbing. The primary-to-secondary leak rate in the ASG is 500 gpd. The primary-to-secondary leak rate path associated with the ASG is direct to the Turbine Building and is terminated at the end of the event.

The ISGs discharge to the environment until RHR is placed in service. The primary-to-secondary leak rate total to the ISGs is 940 gpd, which is the remainder of the modeled 1 gpm (1440 gpd) total accident induced leakage. Due to the effects of partitioning and moisture carryover, the total radionuclides released to the environment are reduced by a factor of 100. The effect of partitioning and moisture carryover is modeled by reducing the steam release rate by a factor of 100 to conserve radionuclides in the ISG liquid. Releases of noble gases are modeled without reduction for partitioning or moisture carryover.

Radionuclides initially in the steam space do not provide any significant dose contribution and are not considered.

There are several nuclide transport models associated with the ASGs and ISGs. The combined results from the cases ensure proper accounting of iodine, particulates, and noble gas releases. Those models are:

1. Release of secondary side bulk liquid that has activity at the TS limit of 0.1 Ci/gm DE I-131. It is assumed that Noble gas is not present in the bulk liquid initial inventory.

2. Release of TS levels of RCS noble gases activity associated with primary-to-secondary leakage.

Serial No.25-017 Docket Nos. 50-280/281 Page 23 of 41

3. Release of TS levels of RCS iodine and particulate activity associated with primary-to-secondary leakage.

4. Release of RCS activity associated with concurrent (500 times the appearance rates generated from RCS activity at 1 Ci/gm DE I-131 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) or pre-accident iodine spiking (10 Ci/gm Dose Equivalent I-131).

4.2.3.2 Faulted MSSV Release The faulted MSSV scenario models the ISGs as described in Section 4.2.3.1. However, the ASG release pathway is not into the Turbine Building, but directly to the environment.

The steam releases from the ASG were increased by a factor of 10 to simulate immediate isolation of AFW and subsequent dry-out of the ASG. No iodine partitioning or moisture carryover was modeled in the ASG. Primary-to-secondary leakage for the ASG is assumed to flash directly to steam and released directly to the environment. There are no releases into the Turbine Building. The models used in this scenario are similar to those documented in Section 4.2.3.1.

4.2.4 MSLB Atmospheric Dispersion Factors The Atmospheric Dispersion Factors (/Q) are unchanged from the CLB submitted in 2018 (References 5 and 6). The EAB location uses the worst 2-hour /Q value for the entire timeline and determines the worst 2-hour dose with a sliding sum. Steam velocities were evaluated for both the MSSV and TDAFW release points. The steam velocities exceed the Table 4-1 minimum steam velocities required for reducing the Control Room X/Qs by a factor of 5.

4.2.5 Key Analysis Assumptions and Inputs 4.2.5.1 Method of Analysis The RADTRAD-NAI code [Reference 3] is used to calculate the radiological consequences from airborne releases resulting from a MSLB at SPS to the EAB, LPZ, and Control Room.

The schematic shown in Figure 4-1 provides an overall picture of the design basis MSLB involving a break into the Turbine Building and releases to the environment. Hot full-power (HFP) mass values were used for secondary side bulk liquid masses consistent with the modeled steam releases.

The evaluation of the break in the Turbine Building considered assumptions which maximize dose to model the release from the ASG into 50% of the Turbine Building volume. High and low escape rates from the louvers, dampers, etc. (i.e., 12 building-volumes/hr down to 0.2 building-volume/hr) are modeled in separate scenarios to maximize resulting offsite and Control Room dose consequences, respectively.

Serial No.25-017 Docket Nos. 50-280/281 Page 24 of 41 Releases from the ASG persist for 30 minutes after event initiation, at which point the steam line break is isolated. Releases from the ISGs persist until the RHR system is placed into service. The time after the event at which the RHR system is put into service and CSD is achieved is 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months (10 days) for the LOOP scenario and 12.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for the no-LOOP scenario. Primary-to-secondary leakage through the ASG is assumed to continue for the duration of the scenarios.

The schematic shown in Figure 4-2 provides a similar picture of the design basis MSLB involving a faulted MSSV. Hot full-power mass values were used for secondary side bulk liquid masses, consistent with the modeled steam releases. There are no releases to the Turbine Building.

Serial No.25-017 Docket Nos. 50-280/281 Page 25 of 41 Figure 4-1: MSLB Turbine Building Release Model Reactor Coolant System 401,790 lbm Turbine Building 3E6 ft3 Affected Steam Generator (ASG) Bulk Liquid 105,700 lbm Credit is not taken for partitioning of iodine and moisture carryover of particulates.

Compartment mass reduced by 50% to aid the release of initial ASG inventory. Initial Ci inventory is conserved by using source term fraction data from Table 4-8.

Intact Steam Generators (ISG)

Liquid - no tube uncovery 105,700 lbm/SG or 211,400 lbm Credit is taken for partitioning of iodine and moisture carryover of particulates.

ASG Primary-to-Secondary leak rate

- 2.90 lbm/min ISG Primary-to-Secondary leak rate - 5.45 lbm/min Exhaust to Environment ISG Bulk Liquid & RCS Activity Release through Safeties - See Table 4-10 Exhaust to Environment Release of Turbine Building Volume through Building Exhaust/Ventilation

- See Table 4-9 Exhaust to Control Room Direct Intake of Turbine Building Volume through Unfiltered Inleakage and Emergency Filtered Ventilation (available after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months through operator action) - See Figure 4-3 TDAFW to Environment ISG Bulk Liquid & RCS Activity Release through TDAFW Exhaust - See Table 4-11 Exhaust to Turbine Building ASG Bulk Liquid Initial Blowdown of Inventory in First 30 Minutes - See Table 4-9 RCS NG & progeny released directly to the ISG in an NG only case without reducing flow for partitioning or moisture carryover. NG case ISG exhaust flows to the environment used 100% efficient filters for P/E/O.

Serial No.25-017 Docket Nos. 50-280/281 Page 26 of 41 Figure 4-2: MSLB Faulted MSSV Release Model Reactor Coolant System 401,790 lbm Affected Steam Generator (ASG) Bulk Liquid 105,700 lbm Credit is not taken for partitioning of iodine and moisture carryover of particulates.

Intact Steam Generators (ISG)

Liquid - no tube uncovery 105,700 lbm/SG or 211,400 lbm Credit is taken for partitioning of iodine and moisture carryover of particulates.

ASG Primary-to-Secondary leak rate to Environment

- 2.90 lbm/min ISG Primary-to-Secondary leak rate - 5.45 lbm/min Exhaust to Environment ISG Bulk Liquid & RCS Activity Release through Safeties -

See Table 4-13 Exhaust to Environment Rapid release of all curie content simulating ASG dry out early in the event. - See Table 4-12 TDAFW to Environment ISG Bulk Liquid & RCS Activity Release through TDAFW Exhaust - See Table 4-14 RCS NG & progeny released directly to the ISG in an NG only case without reducing flow for partitioning or moisture carryover. NG case ISG exhaust flows to the environment used 100% efficient filters for P/E/O.

Serial No.25-017 Docket Nos. 50-280/281 Page 27 of 41 The Control Room modeled in this analysis contains three intake pathways: unfiltered inleakage (UFI), normal ventilation, and emergency ventilation. UFI draws from the environment for the duration of the event at 250 cfm. Normal ventilation operates while power is available, and the Control Room is not isolated at an unfiltered flow rate of 3300 cfm from the environment. Since the LOOP scenario assumes a coincident loss of offsite power, normal ventilation is not modeled in the LOOP cases. The no-LOOP scenario assumes the continued availability of offsite power. Therefore, normal ventilation operates until the Control Room is isolated after the start of the event as a result of a SI signal. The emergency ventilation system draws from the Turbine Building at 900 cfm after operator action places the system into service one hour after event initiation.

Emergency ventilation remains in service for the remainder of the event.

The RADTRAD-NAI code calculates dose consequences based on the nuclide inventories transported between the compartments. The Control Room UFI and emergency intake would draw from the Turbine Building. The Turbine Building volume would be expected to include nuclides from the ASG blow down as well as from the ISG discharge to the environment. To model this scenario, the two sources were modeled as separate nuclide inventories: nuclides from the ASG blowdown in the Turbine Building and nuclides from the ISG discharge to the atmosphere. Thus, the RADTRAD-NAI model includes two UFI intake paths and two emergency intake paths that both use the nominal flow rates for the pathway. While the total flow rates are doubled for each path, the total nuclide inventory transported to the Control Room from each source is conserved. Before isolation with offsite power available (no-LOOP), the Control Room exhaust to the environment is modeled at 3550 cfm (3300 + 250). After isolation with no-LOOP, or for LOOP cases, the Control Room exhaust to the environment is modeled at 250 cfm before one hour and 1150 cfm (900 + 250) after the one hour. Figure 4-3 provides a visual representation of the Control Room model.

Serial No.25-017 Docket Nos. 50-280/281 Page 28 of 41 Figure 4-3: MSLB Control Room Model Turbine Building 3E6 ft3 Control Room 2.23E5 ft3 Emergency Ventilation Filtered intake from Turbine Building 900 cfm, 1 - 720 Hours Emergency Ventilation Filtered intake from Environment 900 cfm, 1 - 720 Hours Unfiltered Inleakage Unfiltered intake from Environment 250 cfm, 0 - 720 Hours Unfiltered Inleakage Unfiltered intake from Turbine Building 250 cfm (after CR Isolation)

Normal Ventilation Unfiltered intake from Environment until CR Isolation (no-LOOP only) 3300 cfm, 0 second - CR isolation Control Room Exhaust Unfiltered release to Environment (equal to the sum of intake flows)

Environment

Serial No.25-017 Docket Nos. 50-280/281 Page 29 of 41 4.2.5.2 Release Rate Data Basic data and assumptions are shown in Table 4-1 and Table 4-2. Additional data include the mass and volumetric flow rates documented below.

Table 4-9: ASG Flow Rates to the Turbine Building and Turbine Building Volumetric Flow Rates to the Environment (Turbine Building Release w/LOOP)

Time (hrs)

Mass Flow Rate to Turbine Building, (lbm/min)

Turbine Building Volumetric Flow Rate to Environment (cfm) 0.0000E+00 0.0000E+00 0.0000E+00 2.7778E-04 2.5741E+05 6.2847E+06 2.7778E-03 1.7035E+05 4.1591E+06 4.4278E-03 1.0032E+05 2.4493E+06 7.5903E-03 6.4001E+04 1.5626E+06 1.1715E-02 4.4355E+04 1.0829E+06 1.7215E-02 3.2636E+04 7.9681E+05 2.5328E-02 2.7484E+04 6.7102E+05 5.0903E-02 1.5667E+04 3.8251E+05 5.3240E-02 3.1883E+03 7.7842E+04 5.0000E-01 0.0000E+00 1.0000E+04 3.0000E+00 0.0000E+00 1.0000E+04 2.4000E+02 0.0000E+00 1.0000E+04 Table 4-10: ISG Flow Rates to the Environment (Turbine Building Release w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 3.7428E-02 5.3143E+03 7.4003E-02 0.0000E+00 5.0000E-01 0.0000E+00 7.1111E-01 1.5611E+03 7.5833E-01 9.1258E+03 7.7361E-01 3.0068E+01

Serial No.25-017 Docket Nos. 50-280/281 Page 30 of 41 Table 4-10: ISG Flow Rates to the Environment (Turbine Building Release w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 8.0000E-01 6.1591E+03 8.2361E-01 4.8678E+01 8.4167E-01 3.8831E+03 9.3333E-01 4.5172E+01 9.5278E-01 7.3960E+03 9.7083E-01 2.9999E+01 9.9722E-01 1.0425E+04 1.0097E+00 0.0000E+00 1.0347E+00 6.2261E+03 1.0556E+00 2.3658E+03 1.1389E+00 7.1509E+03 1.1556E+00 3.9534E+01 1.1778E+00 6.9547E+03 1.1944E+00 2.6477E+03 1.5389E+00 7.6349E+03 1.5514E+00 1.6138E+03 1.6972E+00 9.6853E+03 1.7056E+00 1.1071E+03 2.0042E+00 9.7357E+03 2.0125E+00 5.1547E+02 2.0944E+00 2.3365E+03 2.1472E+00 4.9831E+02 2.2319E+00 9.7297E+03 2.2403E+00 5.0054E+02 2.3264E+00 1.1901E+03 3.0000E+00 1.0199E+03 2.4000E+02 0.0000E+00 Note 1: These Steam Release Rates for all cases except those modeling Noble Gas releases were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the steam generator.

Serial No.25-017 Docket Nos. 50-280/281 Page 31 of 41 Table 4-11: TDAFW Turbine Exhaust Flow Rates (Turbine Building Release w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 2.7778E-04 5.1608E+02 5.5556E-04 5.2717E+02 5.0000E-01 5.2718E+02 3.0000E+00 5.2718E+02 3.9931E+00 5.2694E+02 2.4000E+02 0.0000E+00 Note 1: These Steam Release Rates were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the steam generator.

Table 4-12: ASG Flow Rates to the Environment (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 2.7778E-04 5.7595E+03 6.8615E-02 4.5375E+03 1.4287E-01 3.6846E+03 5.0009E-01 3.8806E+03 8.7222E-01 2.6006E+03 9.1111E-01 1.0918E+03 9.5556E-01 3.4774E+01 3.0000E+00 2.2975E+01 2.4000E+02 0.0000E+00 Note 1: These Steam Release Rates were multiplied by an additional factor of 10 to simulate a rapid dry-out of the steam generator and a rapid release of all radionuclides.

Serial No.25-017 Docket Nos. 50-280/281 Page 32 of 41 Table 4-13: ISG Flow Rates to the Environment (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 7.8333E-03 1.1821E+04 2.8603E-02 0.0000E+00 5.0009E-01 0.0000E+00 9.7222E-01 1.2492E+03 1.0278E+00 2.8295E+03 1.1556E+00 6.0124E+03 1.1778E+00 8.5655E+01 1.2056E+00 8.2204E+03 1.2222E+00 0.0000E+00 1.2500E+00 7.3964E+03 1.2667E+00 2.3664E+03 1.4111E+00 6.8842E+03 1.4278E+00 6.5957E+02 1.4611E+00 9.7749E+03 1.4722E+00 3.1634E+01 1.5000E+00 5.3620E+03 1.5222E+00 2.0169E+03 1.6111E+00 6.7018E+03 1.6278E+00 3.1614E+01 1.6556E+00 9.0258E+03 1.6667E+00 2.4946E+03 1.9222E+00 1.1266E+03 2.0167E+00 8.0642E+03 2.0278E+00 1.6952E+03 2.1611E+00 1.0647E+03 2.9444E+00 2.1358E+03 3.0000E+00 9.5397E+02 2.4000E+02 0.0000E+00 Note 1: Steam Release Rates for all cases except those modeling Noble Gas releases were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the steam generator.

Serial No.25-017 Docket Nos. 50-280/281 Page 33 of 41 Table 4-14: TDAFW Exhaust Flow Rates (Failed MSSV w/LOOP) (see Note 1)

Time (hrs)

Steam Mass Release Rate (lbm/min) 0.0000E+00 0.0000E+00 2.7778E-04 5.2717E+02 5.0009E-01 5.2717E+02 3.0000E+00 5.2718E+02 3.4833E+00 5.2717E+02 2.4000E+02 0.0000E+00 Note 1: Steam Release Rates were divided by an additional factor of 100 to model iodine partitioning and moisture carryover within the steam generator.

4.2.6 Analysis Conservatisms The analysis includes several inputs and assumptions that are conservative and result in retained margin. Examples of these conservatisms are included below.

1) Event Timeline a) The assumed 10-day cooldown duration is beyond the anticipated time to cool down and depressurize the RCS.

b) This timeline was chosen to minimize time-critical operator actions and ensure Operations staff has ample time to recover equipment, choose the optimal long-term recovery strategy, and receive support from plant staff.

2) Dose Consequence Analysis Steam Release Rate (See Figure 4-4 for a visualization of expected conservatism) a) Steam Release input to the dose analyses represent 110% of the thermal hydraulic (T/H) calculation steam releases, and greater than 110% in the ASG to enhance the release of curies.

b) ISG Steam Release rates between 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 10 days are modeled as constant and equal to the ISG release rates corresponding to the core decay heat at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and terminated after 10 days (i.e., no credit for reduction in decay heat over time).

c) From day 7 to 10, use of the RHR system would result in heat removal from the RCS that would also lower the required amount of steaming from the ISGs. RHR heat removal was not credited to maximize steam releases and therefore the predicted dose consequences.

Serial No.25-017 Docket Nos. 50-280/281 Page 34 of 41 Figure 4-4: Steam Release Conservatism Visualization

3) MCR Isolation timing extended beyond the predicted initiating signal and damper closure timing.

4.2.7 Analysis Results The total TEDE to the EAB, LPZ and Control Room from a MSLB is summarized in Table 4-15 for the concurrent and pre-accident spike. The concurrent spike results in the highest dose consequences for both offsite and the Control Room. The limiting case for Control Room doses was a Turbine Building release w/LOOP. The limiting case for offsite doses was the faulted MSSV w/LOOP. The doses are within the acceptance criteria specified in RG 1.183 and 10 CFR 50.67.

Table 4-15: MSLB Accident Dose Summary Accident Location Dose Results (rem TEDE)

Acceptance Criteria (rem TEDE)

Control Room 3.2 5

Serial No.25-017 Docket Nos. 50-280/281 Page 35 of 41 Table 4-15: MSLB Accident Dose Summary Accident Location Dose Results (rem TEDE)

Acceptance Criteria (rem TEDE)

Concurrent Iodine Spike EAB 1.3 2.5 LPZ 0.7 2.5 Pre-Accident Iodine Spike Control Room 2.5 5

EAB 0.3 25 LPZ 0.1 25

5.0 REGULATORY EVALUATION

5.1 APPLICABLE REGULATORY REQUIREMENTS The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50.67, Accident source term applies to all holders of operating licenses issued prior to January 10, 1997, and holders of renewed licenses under part 54 of this chapter whose initial operating license was issued prior to January 10, 1997, who seek to revise the current accident source term used in their design basis radiological analyses. It requires a licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under 10 CFR 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

The NRC may issue the amendment under 10 CFR Part 50.67 only if the applicant's analysis demonstrates with reasonable assurance that:

i.

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

ii.

An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a

Serial No.25-017 Docket Nos. 50-280/281 Page 36 of 41 radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

iii.

Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

The regulations at 10 CFR 50, Appendix A, General Design Criteria for Nuclear Plants, Criterion 19, Control Room, states that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under 10 CFR 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in 10 CFR 50.2 for the duration of the accident.

The proposed change has no effect on the design of the control room or on operator radiation dose, as that protection is provided by other systems required by the Technical Specifications. The proposed change also has no effect on alternate control locations outside of the control room. Therefore, the only aspect of GDC 19 applicable to the proposed change is the criterion to design the control room from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. The proposed change has no effect on the design of the control room and the proposed

Serial No.25-017 Docket Nos. 50-280/281 Page 37 of 41 actions will ensure that the control room temperature is maintained such that the plant may be operated safely from the control room.

Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents At Nuclear Power Reactors, Revision 0, provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. This guide establishes an acceptable alternative source term (AST) and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

5.2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Virginia Electric and Power Company (Dominion Energy Virginia) proposes an update of the Main Steam Line Break (MSLB) Alternative Source Term (AST) dose consequence analysis for Surry Power Station Units 1 and 2. The updated dose analysis was performed in response to industry operating experience and uses a timeline and steam release rates that bound the effect of a Reactor Coolant System (RCS) stagnant loop condition during cooldown to cold shutdown.

Dominion Energy Virginia has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment.

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

Response: No.

The proposed change to update the MSLB AST dose consequence analysis has been analyzed and has been determined to result in acceptable dose consequences, as it continues to meet the dose acceptance criteria specified in 10 CFR 50.67 and RG 1.183. The proposed change will not result in plant operation in a configuration outside the analyses or design basis and does not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition.

Updating the MSLB AST dose consequence analysis does not require any changes to any plant structures, systems, or components (SSCs), has no direct impact upon plant operation or configuration, and does not impact either the initiation of a currently evaluated accident or the mitigation of its consequences.

Serial No.25-017 Docket Nos. 50-280/281 Page 38 of 41 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change will not create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators not previously considered. Other than a change to the AST, there is no significant change to the other parameters within which the plant is normally operated, and no physical plant modifications are being made. Therefore, the possibility of a new or different type of accident is not created.

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

Response: No.

No design basis or safety limits are exceeded or altered by this change. Approved methodologies will be used to ensure the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed changes have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10CFR 50.67 and RG 1.183. The proposed changes will not result in plant operation in a configuration outside the analyses or design basis and do not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, implementation of the proposed license amendment is safe and has no effect on plant operation. The proposed change makes no physical modifications to plant equipment or how equipment is operated or maintained. Consequently, Dominion Energy Virginia concludes the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Serial No.25-017 Docket Nos. 50-280/281 Page 39 of 41 Conclusion Based on the considerations presented above, there is reasonable assurance that: (1) the health and safety of the public will not be endangered by the demonstration that SPS continues to meet applicable design criteria and safety analysis acceptance criteria, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

S Dominion Energy Virginia has reviewed the proposed license amendment for environmental considerations in accordance with 10 CFR 51.22. The proposed license amendment 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 amendment meets the eligibility criterion for categorical exclusion from an environmental assessment as 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 amendment.

7.0 CONCLUSION

S The updated AST dose consequence analysis for the MSLB accident accounts for the effect of loop stagnation and continues to demonstrate the calculated dose results for the Control Room and at the EAB and LPZ meet regulatory requirements.

8.0 COMPUTER FILES An archive of computer files is associated with this License Amendment Request. The archive is comprised of three folders: one folder of RADTRAD input/output files for each of the MSSV w/LOOP and Turbine Building releases w/LOOP accident scenarios, and a folder of RADTRAD input files common to all the RADTRAD analyses. A README file contained within the archive provides additional details. Table 8-1 provides the quality assurance checksum for the data archive.

Table 8-1: Data Archive File Listing Data Archive Filename Checksum Data Files.zip 1823688910

Serial No.25-017 Docket Nos. 50-280/281 Page 40 of 41

7.0 REFERENCES

1.

Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, USNRC, Office of Nuclear Regulatory Research, July 2000.

2.

10 CFR 50.67, Accident Source Term.

3.

Software - RADTRAD-NAI Version 1.3(QA), Numerical Applications Inc.

4.

10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 19 - Control Room (GDC 19).

5.

Letter SN 18-069, Mark D. Sartain to USNRC, Proposed License Amendment Request, Adoption of TSTF-490 and Update of Alternative Source Term Analyses, dated March 2, 2018 (ML18075A021).

6.

Letter SN 19-297, from USNRC to Daniel G. Stoddard, Surry Power Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 295 and 295 to Adopt TSTF-490, Revision 0, and Update Alternative Source Term Analyses (EPID L-2018-LLA-0068), dated June 12, 2019 (ML19028A384).

7.

Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," EPA 520/1-88-020, Environment Protection Agency, 1988.

8.

Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water and Soil," EPA 420-r-93-081, Environmental Protection Agency, 1993.

9.

Callaway LER 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown, (ML18184A389), May 2018.

10. WANO SOER 99-1, Loss of Grid, December 1999, Addendum, December 2004.

11. NUREG-1784, Operating Experience Assessment - Effects of Grid Events on Nuclear Power Plant Performance, December 2003 (ML033530400).

12. IER L2-13-53, Loss of Off-Site Power (LOOP) Analysis, INPO Event Report.

13. INL/EXT-16-39575, Analysis of Loss-of-Offsite-Power Events 1987-2015, Idaho National Laboratory, July 2016.

14. ETE-SU-2017-0069, SPS Emergency Diesel Generator Mission Time, October 2017.

15. NUREG/CR-6890 Vol. 1, Reevaluation of Station Blackout Risk at Nuclear Power Plants, December 2005.

16. INL/RPT-22-68809, Analysis of Loss-of-Offsite-Power Events Update, Idaho National Laboratory, August 2022.

17. ETE-NAF-2023-0067, Dominion Energys NSA Position on Safe-Shutdown (HSD/CSD) Licensing Basis for the VA Units, February 2025.

Serial No.25-017 Docket Nos. 50-280/281 Page 41 of 41

18. NOTEBK-PRA-DOM-HR.3, Revision 4, "Dominion Probabilistic Risk Assessment Model Notebook Human Reliability Analysis Recovery of Loss of Offsite Power,"

November 2015.

19. Letter SN 02-170, USNRC to David A. Christian, "Surry Units 1 and 2 - Issuance of Amendments RE: Alternative Source Term (TAC Nos. MA8649 and MA8650),"

3/8/2002, (ML020710159)

v t e Letter Sequence Request
EPID:L-2025-LLA-0083, (Withdrawn, Closed)
Initiation Acceptance Administration Questions Answers Results Withdrawal Other:
MONTHYEAR2025-12-18 [Table View]

ML25136A297

Text

Enclosures:

1.0 INTRODUCTION AND BACKGROUND

SUMMARY

5.0 REGULATORY EVALUATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

9.0 REFERENCES

1.0 INTRODUCTION AND BACKGROUND

SUMMARY

5.0 REGULATORY EVALUATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

7.0 REFERENCES

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