ML20338A542

Proposed License Amendment Request, Update of the Loss of Coolant Accident Alternate Source Term Dose Analysis

ML20338A542
Person / Time
Site:	Surry
Issue date:	12/03/2020
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
20-381
Download: ML20338A542 (32)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 3, 2020 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST 10 CFR 50.90 Serial No.:

NRA/GDM:

Docket Nos.:

20-381 R1 50-280 50-281 License Nos.: DPR-32 DPR-37 UPDATE OF THE LOSS OF COOLANT ACCIDENT ALTERNATE SOURCE TERM DOSE ANALYSIS Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests amendments to Surry Power Station (Surry) Units 1 and 2 Facility Operating License Numbers DPR-32 and DPR-37, respectively. The proposed License Amendment Request (LAR) updates the Alternate Source Term (AST) analysis for Surry Units 1 and 2 following a Loss of Coolant Accident (LOCA). Specifically, the LOCA AST radiological dose analysis has been revised to increase the assumed containment depressurization profile and to reduce the Refueling Water Storage Tank back leakage limit. A description and assessment of the proposed change is provided in the attachment.

Dominion Energy Virginia has evaluated the proposed amendment and has determined it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in the attachment. We have also determined operation with the proposed change 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 amendment is 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 change.

The LAR has been reviewed and approved by the Facility Safety Review Committee.

Dominion Energy Virginia requests approval of the proposed change by November 30, 2021, with a 90-day implementation period.

Serial No.20-381 Docket Nos. 50-280/281 Page 2 of 3 Should you have any questions or require additional information, please contact Gary D.

Miller at (804) 273-2771.

Respectfully, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support Commitments contained in this letter: None

Attachment:

License Amendment Request - Description and Assessment of the Proposed Changes Associated with the Loss of Coolant Accident (LOCA) Updated Alternate Source Term (AST) Dose Analysis COMMONWEAL TH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President - Nuclear Engineering and 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 ~r"'l day of 3c.c.~, 2020.

My Commission Expires: 0,,...,)V,.bJr ?) \\.J 2.oz..3 GARY DON Notary P Commonwealth Reg.# 76 y Cammi

• 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. Vaughn Thomas NRG Project Manager - Surry Serial No.20-381 Docket Nos. 50-280/281 Page 3 of 3 U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 04 F-12 11555 Rockville Pike Rockville, MD 20852-2738 Mr. G. Edward Miller NRG 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 NRG Senior Resident Inspector Surry Power Station

Attachment LICENSE AMENDMENT REQUEST Serial No.20-381 Docket Nos. 50-280/281 DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGES ASSOCIATED WITH THE LOSS OF COOLANT ACCIDENT {LOCA) UPDATED AL TERNA TE SOURCE TERM {AST) DOSE ANALYSIS Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment LICENSE AMENDMENT REQUEST DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGES ASSOCIATED WITH THE LOSS OF COOLANT ACCIDENT {LOCA) UPDATED AL TERNA TE SOURCE TERM {AST) DOSE ANALYSIS SURRY POWER STATION UNITS 1 AND 2

1.

Introduction and Background 1.1. Introduction The proposed License Amendment Request (LAR) describes the evaluation conducted to assess off-site doses and Control Room habitability at Surry Power Station (SPS)

Units 1 and 2 following a Loss of Coolant Accident (LOCA) per Regulatory Guide (RG) 1.183, Reference [1]. The accident source term discussed in Reference [1] is herein referred to as the Alternative Source Term (AST). The evaluation has employed the detailed methodology contained in Reference [1] 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 guidance in Reference [1].

The application includes the following key elements:

• Increase of the containment depressurization profile, and

• Reduction of the Refueling Water Storage Tank (RWST) back leakage limit.

The revised LOCA radiological dose analysis was performed with a controlled version of the computer code RADTRAD-NAI, Reference [3]. The RADTRAD-NAI computer code calculates the Control Room and offsite doses resulting from releases of radioactive isotopes based on user supplied atmospheric dispersion factors (ADFs), breathing rates, occupancy factors, and dose conversion factors.

Innovative Technology Solutions (ITS) of Albuquerque, New Mexico developed the RADTRAD code for the Nuclear Regulatory Commission (NRC). The original version of the NRC RADTRAD code was documented in NUREG/CR-6604, Reference [4].

The Numerical Applications, Inc. (NAI) version of RADTRAD was originally derived from NRC/ITS RADTRAD, Version 3.01. Subsequently, RADTRAD-NAI was changed to conform to NRC/ITS RADTRAD, Version 3.02, with additional modifications to improve usability.

NAI is currently a division of Zachry Nuclear Engineering, Inc. The RADTRAD-NAI code is maintained under Zachry Nuclear Engineering, lnc.'s Quality Assurance program, which conforms to the requirements of 10 CFR 50, Appendix B.

Control Room ADFs were previously developed using the ARCON96 computer code, Reference [5], following the guidance of RG 1.194, Reference [6]. Control Room ADFs used in this application are unchanged from the values documented in References [16]

Page 1 of 28

and [17].

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Offsite ADFs were previously developed using the PAVAND computer code. PAVAND is Dominion Energy's version of PAVAN V2.0 compiled for use on a Windows PC. The PAVAND code has been used in accordance with NUREG/CR-2858, Reference [12],

and the guidance provided in RG 1.145, Reference [11 ].

Offsite ADFs used in this application are unchanged from the values documented in References [16] and [17].

The core radionuclide inventory for use in determining source term releases was previously generated using the ORIGEN-ARP code from SCALE 4.4a, Reference [7].

The calculations are based on representative design characteristics for the low-leakage cores operated in the SPS units including anticipated core design changes that take advantage of design features of the Westinghouse 15x15 Upgrade fuel with an assumed power level of 2605 MWt.

The power level of 2605 MWt is historically assumed in SPS radiological analyses and slightly exceeds 100.38% of the licensed core rated thermal power (2587 MWt), Reference [8]. The 100.38% corresponds to the power level (2597 MWt), including uncertainties based on the use of ultrasonic flowmeters. The core radionuclide inventory used in this application is unchanged from the values documented in References [16] and [17].

1.2. Current Licensing Basis Summary The current LOCA design basis radiological analysis appears in SPS Updated Final Safety Analysis Report (UFSAR) Section 14.5.5 and was submitted for approval in Reference [16] and was approved in Reference [17]. The analysis was performed using the RADTRAD-NAI code based on a core inventory derived with the ORIGEN-ARP code.

1.3. Analysis Parameters and Assumptions This section describes the general analysis approach and presents analysis parameters, Table 1-1 and Table 1-2, and assumptions, Table 1-3.

The analysis parameters and assumptions presented in Table 1-1, Table 1-2, and Table 1-3 have not changed from the values documented in References [16] and [17].

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 for individuals at the Low Population Zone (LPZ) and for SPS Control Room personnel are calculated for the assumed 30-day duration of the event.

Page 2 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment The TEDE concept is defined to be the Deep Dose Equivalent, DOE, (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 [9], per Section 4.1.2 of Reference [1 ].

The DOE 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 in lieu of ODE in determining the contribution of external dose to the TEDE.

EDE dose conversion factors were taken from Table 111.1 of FGR 12, Reference [1 OJ, per Section 4.1.4 of Reference [1].

Table 1-1: Analysis Parameters Parameter Value Rated Thermal Power 2605 MWt Number of Fuel Assemblies 157 Control Room Filtered Flow 900 cfm (Note 1)

Control Room Filter Efficiency Product Efficiency (%)

Elemental iodine 90 Methyl iodine (organic) 70 Aerosol (particulate) 99 Control Room Volume 223,000 ft3 Control Room Unfiltered lnleakage 250 cfm Control Room Emergency Fan Actuation Time 60 minutes after the start of a LOCA Control Room Isolation Time Prior to beginning of the core inventory release Containment Vacuum Isolation Instantaneous (Note 2)

Containment Spray Flow Rates Table 1-2

• Initiation of ECCS Leakage from the Outside 15 minutes Recirculation Spray (ORS) System Start Time for Crediting Aerosol Removal by 4,100 seconds ( 1.14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />)

Recirculation Spray Outside Recirculation Spray Flow Rates Table 1-2 Inside Recirculation Spray Flow Rates Table 1-2 Containment Operating Deck Elevation 47'4" Page 3 of 28

Parameter Containment Spray Aerosol Removal Coefficients Containment Spray and Recirculation Spray Elemental Iodine Removal Coefficient Containment Spray Elemental Iodine Decontamination Factor (OF)

Time to Achieve Elemental Iodine Decontamination Factor Containment Spray System Wetted Cross Sectional Area Recirculation Spray System Coverage Recirculation Spray Wetted Cross Sectional Area Recirculation Spray Aerosol Removal Coefficients ECCS Leakage (twice the allowable leak rate)

Emergency Control Room Atmospheric Dispersion Factor (Containment releases)

Page 4 of 28 Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Value Time (hr) l\\ (h,1) 0.0278 3.59 0.194 3.69 0.556 4.16 1.0 4.40 1 O h,1 (Note 3) 200 (Note 3) 2.33 hrs (Note 3) 100% (12,500 ft2) 18.78% (One train, Note 4) 37.56% (Both trains) 23.85 %

(23.85% X 12,500 ft2 = 2,980 ft2)

Time (hr) l\\ (h,1) 1.14 30.4 1.8 19.4 1.83 13.5 1.87 7.21 2.02 3.91 2.61 3.43 720.0 0.0 Time (hr)

Filtered Leakage (cc/hr) 0.0 - 0.25 0

0.25-0.5 6,000 0.5-720 6,000 Time (hr)

Unfiltered Leakage (cc/hr) 0.0 -0.5 0

0.5-720 24,000 Time (hrs)

Dispersion Factor (s/m3) 0-2 4.67E-04

Parameter Parameter Emergency Control Room Atmospheric Dispersion Factor (ECCS and RWST releases)

EAB Atmospheric Dispersion Factor LPZ Atmospheric Dispersion Factor Control Room Breathing Rate EAB Breathing Rate LPZ Breathing Rate Control Room Occupancy Factors Core Inventory Dose Conversion Factors Page 5 of 28 Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Value 2-8 3.67E-04 8-24 1.S0E-04 24-96 1.13E-04 96-720 8.78E-05 Value Time (hrs)

Dispersion Factor (s/m3) 0-2 6.55E-04 2-8 4.93E-04 8-24 2.03E-04 24-96 1.44E-04 96-720 1.08E-04 Time (hrs)

Dispersion Factor (s/m3) 0-2 1.02E-03 Time (hrs)

Dispersion Factor (s/m3) 0-8 5.66E-05 8-24 3.84E-05 24-96 1.66E-05 96-720 4.95E-06 Time (hrs)

Breathing Rate (m3/s) 0-720 3.SE-04 Time (hrs)

Breathing Rate (m3/s) 0-720 3.SE-04 Time (hrs)

Breathing Rate (m3/s) 0-8 3.SE-04 8-24 1.8E-04 24-720 2.3E-04 Time (hrs)

Occupancy Factor 0-8 1.0 8-24 1.0 24-96 0.6 96-720 0.4 Table 3-3 References [9] and [1 0]

Parameter Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Value Core Release Fractions, Gap Release Fractions and Release Timing Table 3-1 and Table 3-2 Product Efficiency (%)

Auxiliary Building Filter Efficiency Elemental iodine 90 Methyl iodine (organic) 70 Aerosol (particulate) 99 Product Fractional Distribution Iodine Chemical Form for Containment Release Elemental iodine 0.0485 Methyl iodine (organic) 0.0015 Aerosol (particulate) 0.95 Product Fractional Distribution Iodine Chemical Form for ECCS and RWST Elemental iodine 0.97 Releases Methyl iodine (organic) 0.03 Aerosol (particulate) 0.0 Table 1-1 Notes:

1. The Control Room filtered flow represents the intake of one control room emergency fan (1000 cfm) minus 10% uncertainty (100 cfm).

A lower filtered flow rate is conservative as a higher flow rate leads to dilution of the unfiltered flow to the control room.

2. Containment vacuum isolates within 3.1 seconds of receipt of an isolation signal following the LOCA and is an insignificant contributor to Control Room and offsite dose.

3. The factors used to substantiate the continued use of these parameter values were recalculated based on changes to containment volume, containment sump liquid volume, and spray system coverage in Section 3.4.2.

4. The Recirculation Spray System coverage is the coverage of SPS Unit 2 for one train and is the smaller coverage between both units.

Table 1-2: Spray System Characteristics System / Location Header Elevation Time Flow Rate (gpm) 100 s-700 s 815 Containment Spray / Dome 142' 5" and 143' 9" 700 s-2000 s 850 2000 s - 3600 s 1015 3600 s -4100 s 1090 Page 6 of 28

Containment Spray / Crane 95' 6" Wall Outside Recirculation Spray 93' 5" and 94' 5" Inside Recirculation Spray 93' 5" and 94' 5" Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment 100 s - 700 s 1080 700 s-2000 s 1100 2000 s - 3600 s 1210 3600 s -4100 s 1270 4100 s - 30 days 2900 4100 s - 30 days 2800 Table 1-3: Analysis Assumptions Assumption Description Control Room Dose Contributions from External Negligible (Note 1)

Radiation Sources Available Trains of Containment Spray (CS), Inside One train of each is available and Outside Recirculation Sprays (IRS and ORS)

(Note 2)

Mixing Rate between Sprayed and Unsprayed 2 times the unsprayed volumes Containment Volumes per hour RWST Air Flow to Environment 1000 cfm Table 1-3 Notes:

1. The Control Room walls and ceiling are constructed of at least 2 fl of concrete. The Control Room dose contributions from Containment, Plume, and Filter Shine are assumed to be attenuated to negligible levels because the Control Room walls and roof thickness are at least 46 cm / 18 in of concrete. Reference [13] indicates that 18 inches of concrete is sufficient to attenuate most external design basis accident sources to negligible levels.

2. Only one train of the CS, one train of IRS and one train of ORS were assumed to be working, based on single failure criteria.

2.

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 analyses contained within this attachment. 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 LOCA radiological event are: 1) an increase of the containment depressurization profile, and 2) a reduction of the RWST back leakage Page 7 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment limit.

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 NRG approval.

Additionally, minor changes to model inputs were performed to ensure commonality with the existing containment analyses inputs.

2.1. Increased Containment Depressurization Profile The analysis supports an increase in the containment depressurization profile. The current licensing basis states that the containment will depressurize to 1.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed change would require the containment to depressurize to 2.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This change will be reflected in the LOCA dose consequence analysis through an increased containment leak rate as presented in Table 2-1 and Section 3.5.1.

2.2. Reduced RWST Back Leakage Limit The analysis also supports a decrease in the allowable RWST back leakage.

The current licensing basis employs an allowable RWST back leakage rate of 18,000 cc/hr (analyzed at twice the allowable rate).

The proposed change would reduce the allowable RWST back leakage rate to 9,000 cc/hr (analyzed at twice the allowable rate).

This change is intended to conserve LOCA dose consequence analysis margins.

2.3. Changes to the Design and Licensing Basis Accident Analyses This section provides a comparative summary of the current design and licensing bases (CLB) and the proposed changes discussed in Sections 2.1 and 2.2. The summary is listed in Table 2-1. A detailed discussion of the changes is provided in Section 3.

Table 2-1: Comparison of Proposed Changes Parameter CLB Value Proposed Value Time (hr)

Rate Time (hr)

Rate

(%vol/day)

(%vol/day)

Containment Leak Rate (Note 1) 0.0-1.0 0.1 0.0-1.0 0.1 1.0-4.0 0.029 1.0- 6.0 0.04 4.0- 720 0.0 6.0- 720 0.0 RWST Filtered Back Leakage Time (hr)

Rate Time (hr)

Rate (cc/hr)

(cc/hr)

(Note 2) 0.5-720 36,000 0.5-720 18,000 Containment Volume (Note 3) 1,863,000 ft3 1,819,000 ft3 Page 8 of 28

Parameter Containment Spray System Coverage (Note 4)

Containment Sprayed Volume Spray Unsprayed Volume Coverage Mixing Rate (Note 5)

Recirculation Sprayed Volume Spray Unsprayed Volume Coverage Mixing Rate (Note 5)

Sump Volume (Note 6)

RWST Free Air Volume (Note 7)

CLB Value 60%

1, 118,000 ft3 745,000 ft3 24,840 cfm 349,900 ft3 1,513,000 ft3 50,440 cfm 58,300 ft3 53,350 ft3 Table 2-1 Notes:

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Proposed Value 61%

1,109,590 ft3 709,410 ft3 23,647 cfm 341,608 ft3 1,477,392 ft3 49,246 cfm 55,986 ft3 50,502 ft3

1. The containment leak rate was recalculated as a result of increasing the containment depressurization profile. Additional information is presented in Section 3.5.1.

2. The RWST filtered back leakage represents twice the allowable leak rate, and it was decreased to conserve analysis margin.

3. The containment volume was updated to reflect the value used in containment analyses, Reference [19].

4. The CS System coverage was updated as a result of containment volume change.

5. The CS and RS coverage and mixing rates were updated as a result of containment volume change.

6. The sump volume was updated to use the value calculated by containment analyses.

7. The RWST free air volume was updated based on updated CS pump suction submergence calculations.

3.

Loss of Coolant Accident {LOCA) Dose Reanalysis This section describes the methods employed and results obtained from the radiological reanalysis of the design basis LOCA.

The analysis considers dose from several sources as follows:

Containment Leakage Plume, ECCS Component Leakage, RWST Back Leakage.

Doses are calculated at the EAB for the worst-case two-hour period, at the LPZ Boundary (LPZ), and in the SPS Control Room. The methodology used to evaluate the Page 9 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment doses resulting from a LOCA is consistent with Reference [1 ].

3.1. Scenario Description The design basis LOCA scenario for radiological calculations is initiated assuming a major rupture of the primary Reactor Coolant System (RCS) piping. In order to yield radioactive releases of the magnitude specified in Reference [1], it is also assumed that the ECCS does not provide adequate core cooling, such that significant core melting occurs.

This general scenario does not represent any specific accident sequence; nevertheless, it is representative of a class of severe damage incidents that were evaluated in the development of the Reference [1] source term characteristics. Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the severity of incidents evaluated for design basis transient analysis.

Activity from the core is released to the containment and then to the environment by means of containment leakage and leakage from the ECCS and RWST.

For the containment leakage analysis, the activity released from the fuel is assumed to be in the containment atmosphere until removed by sprays, radioactive decay, or leakage from the containment.

For the ECCS and RWST leakage analysis, the iodine activity released from the fuel is assumed to be in the sump solution until removed by radioactive decay or leakage from the ECCS and RWST.

3.2. Source Term Definition Reference [1] provides explicit description of the key AST characteristics recommended for use in design basis radiological analyses. The core radionuclide inventory used in this analysis is unchanged from the values documented in References [16] and [17] and is provided herein for completeness.

Table 3-1 lists Reference [1] source term inputs used in the LOCA analysis, which includes the core inventory release fractions by radionuclide group, timing of release, and chemical form of the release into containment.

i Reference [1] divides the releases from the core into two phases:

1. The Fuel Gap Release Phase, during the first 30 minutes, and

2. The Early In-vessel Release Phase, during the subsequent 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Table 3-2 shows the fractions of the total core inventory of various isotope groups that are assumed to be released in each of the two phases of the LOCA analysis. Table 3-3 lists the isotopes and the associated curies at the end of a fuel cycle that were input to RADTRAD-NAI. The CEDE and EDE dose conversion factors used for each of the isotopes were based on References [9] and [1 O], respectively.

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Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Table 3-1: Source Term Input Data Characteristic Source Term (based on Reference [1])

Noble Gases 100%

Iodine 40%

Core Fractions Released to Cesium 30%

Containment Tellurium 5%

Barium 2%

Others 0.02% to 0.25%

Timing of Release Released in two phases over 1.8-hour interval Inorganic Vapor 4.85%

Iodine Chemical and Physical Organic Vapor 0.15%

Form Aerosol 95%

Solids Treated as an Aerosol Table 3-2: Release Phases Core Release Fractions (Note 1)

Isotope Group (based on Reference [1])

Gap Early In-Vessel Duration (hours) 0.5 1.3 Noble Gases (Note 2) 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium 0.0 0.05 Barium 0.0 0.02 Strontium 0.0 0.02 Noble Metals 0.0 0.0025 Cerium 0.0 0.0005 Lanthanides 0.0 0.0002 Table 3-2 Notes:

1. Release durations apply only to the Containment release. The ECCS leakage portion of the analysis conservatively assumes that the entire core release fraction is in the Containment sump from the start of the LOCA.

2. Noble gases are not scrubbed from the Containment atmosphere and therefore are not found in either the sump or ECCS fluid.

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Isotope Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-138 Br-82 Br-83 Br-84 1-130 1-131 1-132 1-133 1-134 1-135 Rb-86 Rb-88 Rb-89 Cs-134 Cs-134m Cs-136 Cs-137 Cs-138 Sb-125 Sb-126 Sb-127 Sb-129 Te-125m Te-127 Te-127m Te-129 Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Table 3-3: Core Inventory Isotope Group Curies Noble gas 8.662E+06 Noble gas 7.619E+05 Noble gas 1.814E+07 Noble gas 3.647E+07 Noble gas 5.072E+07 Noble gas 9.262E+05 Noble gas 1.375E+08 Noble gas 4.226E+06 Noble gas 4.326E+07 Noble gas 3.007E+07 Noble gas 1.237E+08 Halogen 2.227E+05 Halogen 8.663E+06 Halogen 1.620E+07 Halogen 1.537E+06 Halogen 6.949E+07 Halogen 1.023E+08 Halogen 1.442E+08 Halogen 1.602E+08 Halogen 1.372E+08 Alkali Metal 1.224E+05 Alkali Metal 5.199E+07 Alkali Metal 6.797E+07 Alkali Metal 1.189E+07 Alkali Metal 2.749E+06 Alkali Metal 3.749E+06 Alkali Metal 8.820E+06 Alkali Metal 1.337E+08 Tellurium 5.697E+05 Tellurium 2.714E+04 Tellurium 6.100E+06 Tellurium 2.264E+07 Tellurium 1.212E+05 Tellurium 6.033E+06 Tellurium 9.847E+05 Tellurium 2.147E+07 Page 12 of 28

Isotope Te-129m Te-131 Te-131m Te-132 Te-133 Sr-89 Sr-90 Sr-91 Sr-92 Ba-137m Ba-139 Ba-140 Ba-141 Pd-109 Ag-111 Mo-99 Rh-103m Rh-105 Rh-106 Ru-103 Ru-105 Ru-106 Tc-99 Tc-99m Tc-101 Ce-141 Ce-143 Ce-144 Np-239 Pu-238 Pu-239 Pu-240 Pu-241 Am-241 Cm-242 Cm-244 Eu-156 La-140 La-141 Isotope Group Tellurium Tellurium Tellurium Tellurium Tellurium Barium-Strontium Barium-Strontium Barium-Strontium Barium-Strontium Barium-Strontium Barium-Strontium Barium-Strontium Barium-Strontium Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Noble Metal Cerium Cerium Cerium Cerium Cerium Cerium Cerium Cerium Lanthanides Lanthanides Lanthanides Lanthanides Lanthanides Lanthanides Page 13 of 28 Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Curies 4.310E+06 5.902E+07 1.403E+07 1.003E+08 7.840E+07 7.078E+07 6.646E+06 8.839E+07 9.341E+07 8.378E+06 1.278E+08 1.277E+08 1.158E+08 2.651E+07 4.266E+06 1.313E+08 1.104E+08 7.091E+07 4.213E+07 1.105E+08 7.723E+07 3.774E+07 1.112E+03 1.164E+08 1.192E+08 1.172E+08 1.100E+08 9.043E+07 1.471 E+09 2.456E+05 2.549E+04 3.477E+04 9.362E+06 1.051 E+04 3.000E+06 2.907E+05 1.562E+07 1.387E+08 1.168E+08

Isotope Isotope Group La-142 Lanthanides La-143 Lanthanides Nb-95 Lanthanides Nb-95m Lanthanides Nb-97 Lanthanides Nb-97m Lanthanides Nd-147 Lanthanides Pm-147 Lanthanides Pm-148 Lanthanides Pm-148m Lanthanides Pm-149 Lanthanides Pr-143 Lanthanides Pr-144 Lanthanides Pr-144m Lanthanides Sm-153 Lanthanides Y-90 Lanthanides Y-91 Lanthanides Y-91m Lanthanides Y-92 Lanthanides Y-93 Lanthanides Y-94 Lanthanides Y-95 Lanthanides Zr-95 Lanthanides Zr-97 Lanthanides Sb-124 Tellurium Te-133m Tellurium Te-134 Tellurium Eu-154 Lanthanides Eu-155 Lanthanides Pm-151 Lanthanides Am-242 Lanthanides Np-238 Cerium Pu-243 Cerium 3.3. Atmospheric Dispersion Factors {X/Qs)

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Curies 1.143E+08 1.091 E+08 1.212E+08 1.341E+06 1.148E+08 1.083E+08 4.704E+07 1.334E+07 1.176E+07 2.171E+06 4.151E+07 1.062E+08 9.102E+07 1.270E+06 3.093E+07 6.858E+06 9.119E+07 5.127E+07 9.411E+07 7.117E+07 1.125E+08 1.168E+08 1.206E+08 1.140E+08 5.145E+04 6.459E+07 1.287E+08 5.029E+05 2.056E+05 1.408E+07 5.259E+06 2.755E+07 2.721E+07 The Control Room atmospheric dispersion factors (X/Qs), listed in Table 1-1, are unchanged from the values documented in References [16] and [17]. The EAB and LPZ X/Qs, listed in Table 1-1, are unchanged from the values documented in Page 14 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment References [16] and [17]. The EAB and LPZ X/Q values were conservatively modeled using a ground-level release without credit for building wake effects.

3.4.

Containment Spray (CS)

The current licensing basis for the LOCA uses sprays in containment to remove elemental and particulate iodine from the containment atmosphere.

There are two spray systems in containment; 1) the CS system, which uses water from the RWST, and 2) the Recirculation Spray (RS) system, which uses recirculated water from the containment sump.

The CS system has two trains with three 360-degree ring headers, two at the dome and one at the top of the crane wall (the crane wall header is fed from both trains). The RS system consists of two trains (inside RS and outside RS) with two subsystems each for a total of four RS subsystems. Each subsystem consists of a RS pump, a RS heat exchanger and 180-degree spray ring header for a total of four 180-degree crane wall ring headers. The spray ring headers are located approximately 4 7 feet above the operating floor in containment. The RS system design is such that loss of one train reduces spray coverage to 180 degrees.

One train of the CS system and one train of the RS system are assumed to operate following the LOCA. The CS system start is based on CLS High-High pressure. The RS system start is based on a coincident CLS High-High pressure signal and RWST Level Low signal, which is assumed to occur at 0.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to account for the RS system contribution to ECCS leakage. However, RS is not credited for iodine removal until the CS system stops. This is conservative as there is no overlap of spray systems. The analysis conservatively assumes that CS is terminated 1.14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the start of the event.

3.4.1. Containment Spray Removal of Aerosol (Particulates)

The calculation of CS aerosol removal coefficients follows the method described in Reference [15]. The inputs to the methodology include the fall height "H" of the water droplets, in centimeters, and the spray water flux "Q", in cm/s. The water droplets' fall heights for the CS and RS systems, calculated as the difference between spray system header elevation, Table 1-2, and the elevation of containment operations deck, remain unchanged from the values documented in References [16] and [17]. The CS and RS systems' flow rates, Table 1-2, are unchanged from the values documented in References [16] and [17].

Thus, the CS and RS aerosol removal coefficients from Reference [16], presented in Table 1-1, have been used in this analysis.

3.4.2. Containment Spray Removal of Elemental Iodine Reference [14], Section 6.5.2, identifies a methodology previously used and approved Page 15 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment for the determination of spray removal of elemental iodine. The elemental iodine spray removal coefficients are calculated using Equation 3-1.

I _

6 x Kg x T x F

,15 -

V x D X 3600 hr I

Equation 3-1 Where:

T=

T=

T=

F=

K9 = gas-phase mass transfer coefficient, 7.5 cm/s, T = drop fall time as ratio of fall height to velocity for a drop terminal velocity of 400

emfs, Elevation CS Dome Headers: Table 1-2 and Equation 3-2, Elevation CS Crane Wall Headers: Table 1-2 and Equation 3-3, Elevation RS Headers: Table 1-2 and Equation 3-4, Elevation operating deck: Table 1-1 F = spray volumetric flow rate, cm3/s, Table 1-2 and Equation 3-5, V = containment building net free volume, cm3, Table 1-1, Table 1-2 and Equation 3-6, D = mean spray drop diameter, 1000 µm.

H 142 15 - 47'4 95 11 ft 2900 cm

=

cm = 7.25 s Equation 3-2 400cm -

400cm

- 400cm 400-s s

s s

H 95 16 - 47'4 48 12 ft 1470 cm cm = 3.675 s Equation 3-3 400cm -

400cm - 400cm - 400-s s

s s

H 93 15-47'4 46 11 ft 1400 cm

=

=

cm = 3.5 s Equation 3-4 400cm -

400cm 400cm 400-s s

s s

(Flow Rate gpm) X ( 0.13368 ~~~) X ( (30.48) 31n;:)

Equation 3-5 60~

mm V = (Free Containment Volume) x (Spray System Coverage)

(

cm')

X (30.48) 3 f t 3 Equation 3-6 The elemental iodine spray removal coefficients for the CS and RS system are presented in Table 3-4. These elemental iodine spray removal coefficient calculations substantiate the continued use of an elemental iodine spray removal coefficient of 10 hr1.

Page 16 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Table 3-4: CS and RS Elemental Iodine Removal Coefficients F

V D

System I Location Spray Volumetric Containment Mean Spray J!.5 (hr-1)

Drop Diameter Flow Rate (cm3/s)

Volume (cm3)

(cm)

CS Dome Headers 51,418.3 3.142E+10 0.1 19.2 CS Crane Wall 68,137.1 3.142E+10 0.1 12.9 Headers CS Total 32.1 RS 359,612.6 9.673E+09 0.1 210.8 Reference [14], Section 6.5.2, also provides the equation for maximum elemental iodine decontamination factor (OF) as:

I v; X H DF = 1 + s \\1c I Equation 3-7 Where:

H = effective iodine partition coefficient, 5000, Vs= volume of liquid in containment sump, 55,986 ft3, Table 2-1, Ve= containment building net free volume less Vs, ft3*

After CS termination at 1.14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, the containment volume being sprayed by one train of RS is 18. 78%, Table 1-1. The OF cutoff is applied after the end of the early in-vessel period of 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and is applied only to 18. 78% of the containment volume.

Consequently, the Ve term in Equation 3-7 is corrected by the fraction of containment volume that is sprayed. The OF is calculated using Equation 3-7 as follows:

55986ft3 X 5000 DF = 1 +--------- = 846 4 0.1878 X (1819000 - 55986)

Assuming full safeguards response, the containment volume sprayed by both trains of RS is 37.56%, Table 1-1.

In this case, the OF is calculated using Equation 3-7 as follows:

55986ft3 X 5000 DF = 1 +--------- = 423 7 0.3756 X (1819000 - 55986)

The calculated maximum elemental iodine decontamination factors substantiate the continued use of the maximum allowable OF of 200.

As prescribed in Reference [14], Section 6.5.2, the removal of elemental iodine by Page 17 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment sprays continues at a rate of 10 h,1 until a OF of 200 is reached. This OF is reached when the elemental iodine activity in the containment at the end of the early in-vessel release phase is reduced by a factor of 200. The time interval it takes to reduce the activity by 200 is calculated using Equation 3-8.

A0 ln(DF) ln(200)

A = A 0 x e-;txt DF = - = e;txt t = --- t = ---

A J

10 Equation 3-8

= 0.53 hr This is the duration required starting at the end of early in-vessel release phase at 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Therefore, the post-accident time at which elemental iodine removal stops is 2.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />.

3.5.

Analysis Methods and Models The RADTRAD-NAI code, Reference [3], is used to calculate the radiological consequences from airborne releases to the EAB, LPZ, and Control Room resulting from a LOCA at SPS.

The RADTRAD-NAI models used in this calculation include the following pathways:

Activity from the failed fuel enters the containment and is released to the atmosphere through containment leakage. The nuclides are released through this pathway. This pathway is not filtered.

Negative pressure in the Containment Building is established within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the accident, Table 1-3. During the first 6-hour interval, no credit is taken for filtering the Containment Building leak rate.

Activity in the sump leaks out of containment via the ECCS and is released to the Safeguards and Auxiliary Buildings and then to the environment. Only iodine is released through this pathway. The release pathway from Safeguards to the environment is modeled as filtered.

The release pathway from the Auxiliary Building to the environment via the Safety Injection (SI) and Charging System is modeled as not filtered.

Modeling the release pathways in this manner is conservative relative to the plant configuration during the postulated LOCA.

Activity in the ECCS leaks back to the RWST.

The RWST back leakage is modeled as a filtered release from Safeguards.

3.5.1. Containment Leakage Model An updated SPS containment depressurization profile is evaluated in this analysis, as discussed in Section 2.1.

Reference [1 ], Appendix A, position 3. 7, states that the primary containment should be assumed to leak at the peak pressure Technical Specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage from sub-atmospheric containments is assumed to terminate when the containment is brought to and Page 18 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment maintained at a sub-atmospheric condition as defined by the Technical Specifications (TS). The SPS TS state that the containment leakage rate acceptance criterion is less than or equal to 0.1 % per day at calculated peak pressure.

The TS maximum containment pressure is 45 psig.

Between O and 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after event initiation, the containment leak rate is 0.1 %

containment volume per day. The volumetric leak rate corresponding to a containment leak rate of 0.1 % containment volume per day at the containment peak pressure of 45 psig is calculated using Equation 3-9.

0.1% 1 f 3 100o/c d x 1,819,000 t Volumetric Leak Rate45 psig =

0 ay

= 1.263 cfm 1440mm day Equation 3-9 Between 1 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after event initiation, the containment volumetric leak rate as a function of containment pressure was modeled as compressible flow through an orifice sized to allow a flow equal to the design leak rate of 0.1 % of containment volume per day at a pressure of 45 psig. The containment volumetric leak rates calculated based on this approach were found to be conservative and acceptable, Section 2.1.1 of Reference [18].

The volumetric leak rate at two different containment pressures is calculated using Equation 3-10. Equation 3-10 relies on conservative key inputs to obtain a conservative containment volumetric leak rate for pressures less that the TS maximum containment pressure, as follows:

Containment design leak rate at 45 psig, Constant containment atmosphere temperature at the design value (280°F) as the containment pressure decreases, Orifice configuration versus a diffuse "area source" for containment releases.

I Equation 3-1 0 Where:

q2 = containment volumetric leak rate at any pressure between 45 psig and 0.1 psig, cfm, q1 = containment volumetric leak rate at 45 psig, calculated using Equation 3-9,

cfm,

!::.P2 = selected pressure, psig, p2 = containment air density at the selected pressure, 280°F, lbm/ft3,

!::.P1 = TS containment maximum pressure, 45 psig, and Page 19 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment p1 = containment air density at the TS containment maximum pressure of 45 psig, 280°F, lbm/ft3.

The containment air density is calculated using Equation 3-11.

(

p + 14.7)

(

560 )

p = 0.0709 X 14.7 X T + 460 Equation 3-11 Where:

P = containment atmosphere pressure, psig, T = containment atmosphere temperature, °F.

Based on the design volumetric leak rate, Equation 3-9, and the updated containment volume, Table 2-1, Equation 3-10 is used to calculate the containment volumetric leak rates as a function of containment pressure. The results are presented in Table 3-5.

Table 3-5: Containment Volumetric Leak Rate as Function of Containment Pressure Containment Pressure Containment Volumetric Leak Rate (psig)

(cfm) 0.0 0.0 0.1 0.120 0.2 0.169 0.3 0.206 0.4 0.237 0.5 0.264 0.6 0.288 0.7 0.310 0.8 0.331 0.9 0.349 1.0 0.367 2.0 0.504 3.0 0.599 4.0 0.673 5.0 0.733 10.0 0.926 15.0 1.034 20.0 1.105 25.0 1.155 Page 20 of 28

Containment Pressure (psig) 30.0 35.0 40.0 45.0 Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Containment Volumetric Leak Rate (cfm) 1.192 1.221 1.244 1.263 Between 1 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after event initiation, a containment pressure of 2 psig generates a volumetric leak rate of 0.504 cfm, Table 3-5. The containment leak rate corresponding to a volumetric leak rate of 0.504 cfm is calculated using Equation 3-12.

0.504 Leak Rate2 psig = -1 6 x 0.1 % = 0.04 % containment volume per day

.2 3 Equation 3-12 The containment leak rate reflecting the updated SPS containment depressurization profile is documented in Table 2-1.

The SPS containment depressurization profile returns to sub-atmospheric conditions 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after event initiation.

Figure 3-1 graphically shows the model used to analyze the containment release for SPS LOCA.

Figure 3-1: Containment Release for the LOCA Analysis Model Sprayed Unfiltered....

Region Intake t

t Unfiltered Filtered

~

Control Release Environment Intake Room Paths Unsprayed CR Region

'Discharge 3.5.2. ECCS Leakage Model The ECCS fluid consists of the contaminated water in the containment sump. Forty percent of the core inventory of iodine isotopes was conservatively modeled as being instantaneously transported from the core to the containment sump.

The fractional distribution of iodine chemical forms is modeled in accordance with Reference [1], as documented in Table 1-1. During a LOCA, the highly radioactive fluid is pumped from the containment sump to the RS headers and sprayed back into the containment sump.

Also, following a design basis LOCA, valve realignment occurs to switch the suction Page 21 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment water source for the ECCS pumps from the RWST to the containment sump.

ECCS leakage develops when Engineered Safeguards Feature (ESF) systems circulate sump water outside containment and leaks develop through packing glands, pump shaft seals and flanged connections. In accordance with Reference [1], the ECCS analysis makes use of two times the sum of the simultaneous leakage from the components in the ESF recirculation systems. The leakage of recirculating sump fluids start time along with filtered and un-filtered leakage rates are listed in Table 1-1. The ORS portion of ECCS leakage to the environment was modeled with 10% iodine evolution as a 90%

efficient filter.

According to Reference [1], when the temperature of the ECCS leakage is less than 212°F or the calculated flash fraction is less than 10%, the amount of iodine that becomes airborne is 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.

The calculated flashing fraction for ECCS fluid using the constant enthalpy process described in Reference [1] is less than 10%; therefore, the flashing fraction is assumed tobe10%.

Figure 3-2 graphically shows the model used to analyze the ECCS release for SPS LOCA.

Figure 3-2: ECCS Release for the LOCA Analysis Model Filtered Release Path...

Unfiltered...

(Safeguards) '

Intake Sump Environment Filtered....

Control Intake Room Unfiltered RP.IP.R~P. P::ith...

~

CR

~ Discharge Auxiliary Building) 3.5.3. ECCS Back Leakage to RWST Model The current licensing basis employs an allowable RWST back leakage rate of 18,000 cc/hr (analyzed at twice the allowable rate).

The analysis presented herein evaluates the LOCA dose consequences based on an allowable RWST back leakage rate of 9,000 cc/hr. This change is intended to conserve LOCA dose consequence analysis margins while providing 50% margin relative to the Surry RWST back leakage testing acceptance criterion.

Page 22 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Figure 3-3 shows graphically the model used to analyze the RWST back leakage release for SPS LOCA.

Figure 3-3: RWST Release for the LOCA Analysis Model Sump Unfiltered....

Intake,

i Environment Filtered

~

Control Intake Room Filtered CR RWST Release Path '

'Discharge 3.5.4. Control Room Model The Control Room volume is listed in Table 1-1. The LOCA causes an SI signal, which also isolates the Control Room (per current Licensing Basis). The Control Room is isolated within 20 seconds after the SI signal. Based on Reference [1 ], the onset of the gap release does not start until 30 seconds post-LOCA. Therefore, the Control Room will be isolated prior to the arrival of the radioactive release.

Control Room parameters are provided in Table 1-1. These parameters include the normal operation in-leakage, the emergency operation flow rates, Control Room volume, filter efficiencies and Control Room operator breathing rates. The atmospheric dispersion factors in Table 1-1 are calculated for release of activity from the release point to the point of intake to the Control Room to determine the activity available for intake. The inflow to the Control Room and the Control Room recirculation flow are used to calculate the activity introduced to the Control Room and cleanup of activity from that flow. The Control Room filter efficiencies are provided in Table 1-1.

The post-LOCA dose consequences to the SPS Control Room are due to the following sources:

Containment leakage, ECCS leakage, and RWST leakage.

3.6.

Results The calculated LOCA radiological consequences are compared with the limits provided in Reference [2].

Page 23 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment Dose calculations are performed at the EAB for the worst 2-hour period, and for the LPZ and the Control Room for the duration of the accident (30 days). The TEDE dose to the EAB, LPZ and Control Room from a LOCA at SPS are provided in Table 3-6. The LOCA radiological consequences for SPS meet the limits specified in Reference [2],

Section (b )(2), and Reference [1].

Table 3-6: LOCA Dose Acceptance Criteria Location TEDE (rem)

Limits (rem)

EAB 10.6 25.0 LPZ 2.1 25.0 Control Room 4.7 5.0

4.

Conclusions The proposed changes to the containment depressurization profile and the allowable RWST back leakage have been incorporated into the reanalysis of LOCA radiological effects for SPS Units 1 and 2. The analysis results meet the acceptance criteria as specified in References [1] and [2].

5.

References

1.

Regulatory Guide 1.183, "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 1.2(QA), Numerical Applications Inc.

4.

NUREG/CR-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," USNRC, June 1997, S.L. Humphreys et al.

5.

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

6.

Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, June 2003.

Page 24 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment

7.

Software - SCALE Version 4.4a, Oak Ridge National Laboratory, NUREG/CR-0200, Rev. 6, USNRC, May 2000.

8.

Letter from L. N. Hartz of Virginia Electric and Power Company (Serial No.09-223) to USNRC dated January 27, 2010, "License Amendment Request, Measurement Uncertainty Recapture Power Uprate," (ML100320264).

9.

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.

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

11. Regulatory Guide 1.145, Rev. 1, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," November 1982 (Reissued February 1983).

12. NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Plants," (November 1982).

13. NUREG-0800, "Standard Review Plan," Section 6.4, "Control Room Habitability System," Revision 2, July 1981.

14. NUREG-0800, "Standard Review Plan," Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," U.S. Nuclear Regulatory Commission, Revision 2, December 1988.

15. NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays," June 1993.

16. Letter from Mark D. Sartain of Virginia Electric and Power Company to USNRC (Serial No.18-069), dated March 2, 2018, "Virginia Electric and Power Company, Surry Power Station Units 1 And 2, Proposed License Amendment Request, Adoption of TSTF-490 and Update of Alternative Source Term Analyses,"

(ML18075A021 ).

17. Letter from USNRC to Daniel G. Stoddard of Virginia Electric and Power Company dated June 12, 2019, "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)," (ML19028A384).

Page 25 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment

18. Letter from USNRC to David A. Christian of Virginia Electric and Power Company dated March 8, 2002, "Surry Units 1 and 2 -

Issuance of Amendments Re:

Alternative Source Term (TAC Nos. MA8649 and MA8650)" (ML020710159).

19. DOM-NAF-3-P-A, Gothic Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment.

6.

No Significant Hazards Consideration Determination Virginia Electric and Power Company (Dominion Energy Virginia) is providing a license amendment request to revise the Loss of Coolant Accident (LOCA) Alternate Source Term (AST) dose consequences analysis for Surry Power Station Units 1 and 2. The proposed update of the LOCA AST dose consequences analysis would implement the following changes: (1) increase the containment depressurization profile during a LOCA, and (2) reduction of the Refueling Water Storage Tank (RWST) back leakage limit.

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 for radiological events related to an increase of the containment depressurization profile and a reduction of the RWST back leakage limit has been analyzed and results in acceptable consequences, meeting the criteria as specified in 10 CFR 50.67 and RG 1.183 for the Exclusion Area Boundary (EAB),

Low Population Zone (LPZ), and Control Room doses. 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 i~ a safe operating condition.

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

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

Response: No.

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Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment 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; thus, 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.

Therefore, the changes do not involve a significant reduction in a margin of safety.

Based on the above, implementation of the proposed license amendment is safe and will have no effect on plant operation. The proposed change will make no physical modifications to equipment or how equipment is operated or maintained.

Dominion Energy Virginia concludes that the proposed changes present 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.

7.

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 Commission's 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.

8.

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 Page 27 of 28

Serial No.20-381 Docket Nos. 50-280/281 Revised LOCA AST Dose Analysis Attachment 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.

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