Text

License Amendment Request to Modify the Salem and Hope Creek Exclusion Area Boundary
	ML23249A261
Person / Time
Site:	Salem, Hope Creek
Issue date:	09/06/2023
From:	Mannai D; Public Service Enterprise Group
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23249A260	List: LR-N23-0050, Attachment 5: Hope Creek Pavan Output File; ML23249A261; ML23249A262; ML23249A263; ML23249A264; ML23249A265; ML23249A266; ML23249A267; ML23249A268;
References
	LR-N23-0050, LAR S23-04, LAR H23-02
	Download: ML23249A261 (1)
	v • d • e

PSEG Nuclear LLC P.O . Box 236 , Hancocks Bridge, New Jersey 08038-0236 0 PSEG NuclearLLC 10 CFR 50.90 LR-N23-0050 LAR S23-04 LAR H23-02 September 6, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

License Amendment Request to Modify the Salem and Hope Creek Exclusion Area Boundary In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) requests amendments to Renewed Facility Operating License Nos. DPR-70 and DPR-75 for Salem Generating Station Units 1 and 2 (Salem) and NPF-57 for Hope Creek Generating Station (Hope Creek). The proposed amendments would change the licensing basis as described in the Salem and Hope Creek Updated Final Safety Analysis Reports (UFSARs) to account for modifications to the Exclusion Area Boundary (EAB) for Salem and Hope Creek.

Land parcels designated for use by the New Jersey Wind Port (NJWP) project will be removed from the exclusion area. Although the updated dose consequences for all analyses remain within the limits of 10 CFR 50.67, NRC approval is required for those analyses for which the increase in consequences is more than minimal or that result in a departure from a method of evaluation described in the UFSAR, in accordance with 10 CFR 50.59(c)(2).

The Enclosure to this letter provides a detailed description and evaluation of the proposed changes. Attachment 1 contains the hourly meteorological data for 2019 through 2021.

Attachments 2 and 3 contain the printed input and output PAVAN text files for Salem.

Attachments 4 and 5 contain the same information for Hope Creek.

[DATE] 10 CFR 50.90 Page 2 LR-N23-0050 PSEG requests review and approval of this license amendment request within one year of acceptance. Once approved, the amendment shall be implemented within 180 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If there are any questions or if additional information is needed, please contact Mr. Michael Wiwel at 856-339-7907.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 09/06/2023 (Date)

Respectfully, Digitally signed by Mannai, Mannai, David David Date: 2023.09.06 13:27:40 -04'00' David J Mannai Executive Director - Regulatory Affairs and Nuclear Oversight PSEG Nuclear

Enclosure:

Evaluation of the Proposed Changes Attachments: 1. PSEG Site Hourly Meteorological Data (2019 - 2021)

2. Salem PAVAN Input files

3. Salem PAVAN Output files

4. Hope Creek PAVAN Input files

5. Hope Creek PAVAN Output files cc: Administrator, Region I, NRC Mr. J. Kim, NRC Project Manager, NRC NRC Senior Resident Inspector, Salem NRC Senior Resident Inspector, Hope Creek Ms. A. Pfaff, Manager, NJBNE PSEG Corporate Commitment Tracking Coordinator Site Commitment Tracking Coordinator

LR-N23-0050 LAR S23-04 LAR H23-02 Enclosure Evaluation of the Proposed Change

Subject:

License Amendment Request to Modify the Salem and Hope Creek Exclusion Area Boundary Contents 1

SUMMARY

DESCRIPTION ................................................................................................... 1 2 DETAILED DESCRIPTION .................................................................................................... 1 2.1 Background .................................................................................................................... 1 2.2 Current Salem and Hope Creek Exclusion Areas .......................................................... 1 2.3 Salem Current Design Basis Radiological Analyses Summary ...................................... 3 2.4 Hope Creek Current Design Basis Radiological Analyses Summary ............................. 3 2.5 Reason for the Proposed Change .................................................................................. 3 2.6 Description of the Proposed Changes ............................................................................ 4 2.6.1 Modified Exclusion Area Boundary ......................................................................... 4 2.6.2 Changes to Design Basis Radiological Consequence Analyses ............................. 7 3 TECHNICAL EVALUATION ................................................................................................... 9 3.1 Determination of Atmospheric Dispersion Factors (/Q) ................................................ 9 3.1.1 Meteorological Data ................................................................................................ 9 3.1.2 EAB /Q................................................................................................................. 13 3.2 Salem Loss of Coolant Accident (LOCA) ..................................................................... 21 3.2.1 LOCA Scenario Description .................................................................................. 21 3.2.2 LOCA Source Term Definition ............................................................................... 21 3.2.3 LOCA Atmospheric Dispersion Factors ................................................................. 23 3.2.4 Containment Leakage ........................................................................................... 23 3.2.5 Engineered Safety Feature (ESF) System Leakage ............................................. 25 3.2.6 Containment Vacuum Relief Line Release ............................................................ 26 3.2.7 Post-LOCA Back-leakage to the Refueling Water Storage Tank (RWST) ............ 26 3.2.8 Data for LOCA Model ............................................................................................ 26 3.2.9 Salem LOCA Results ............................................................................................ 29 3.3 Salem Steam Generator Tube Rupture (SGTR) Accident ............................................ 30 3.3.1 SGTR Scenario Description .................................................................................. 30 3.3.2 SGTR Source Term ............................................................................................... 32 3.3.3 SGTR Release Transport ...................................................................................... 33 3.3.4 SGTR Atmospheric Dispersion Factors ................................................................ 35

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LR-N23-0050 LAR S23-04 LAR H23-02 3.3.5 SGTR Key Analysis Assumptions and Inputs ....................................................... 35 3.3.6 SGTR Analysis Results ......................................................................................... 37 3.4 Other Affected Salem Analyses .................................................................................... 40 3.5 Hope Creek Loss of Coolant Accident (LOCA) ............................................................ 41 3.5.1 LOCA Scenario Description .................................................................................. 41 3.5.2 LOCA Source Term Definition ............................................................................... 41 3.5.3 LOCA Atmospheric Dispersion Factors ................................................................. 44 3.5.4 Containment Activity and Leakage ........................................................................ 44 3.5.5 Containment Leakage ........................................................................................... 45 3.5.6 Engineered Safety Feature (ESF) System Leakage ............................................. 46 3.5.7 Post-LOCA MSIV Leakage Pathway ..................................................................... 47 3.5.8 Data for LOCA Model ............................................................................................ 48 3.5.9 Hope Creek LOCA Results ................................................................................... 50 3.6 Hope Creek Main Steam Line Break ............................................................................ 52 3.6.1 MSLB Scenario Description .................................................................................. 52 3.6.2 MSLB Source Term Definition ............................................................................... 52 3.6.3 MSLB Atmospheric Dispersion Factors ................................................................. 53 3.6.4 MSLB Key Analysis Assumptions and Inputs ........................................................ 53 3.6.5 MSLB Analysis Results ......................................................................................... 55 3.7 Other Affected Hope Creek Analyses ........................................................................... 56 4 REGULATORY EVALUATION ............................................................................................ 57 4.1 Applicable Regulatory Requirements/Criteria ............................................................... 57 4.2 No Significant Hazards Consideration Determination Analysis .................................... 57 4.3 Conclusions .................................................................................................................. 59 5 ENVIRONMENTAL CONSIDERATION ............................................................................... 59 6 REFERENCES .................................................................................................................... 60

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LR-N23-0050 LAR S23-04 LAR H23-02 List of Tables Table 2.3-1 Current Salem Design Basis Radiological Analyses ................................................. 3 Table 2.4-1 Current Hope Creek Design Basis Radiological Analyses ........................................ 3 Table 2.6-1 Minimum Exclusion Area Boundary Distances .......................................................... 7 Table 3.1-1 Pasquill Stability Class A - G Distribution of Occurrences by Speed and Direction. 11 Table 3.1-2 Frequency of Occurrence (% of Total Observations) (Calms Distributed In Lowest WS Class) ................................................................................................................................... 12 Table 3.1-3 Hope Creek Summary of Common PAVAN Inputs .................................................. 18 Table 3.1-4 Salem Summary of Common PAVAN Inputs ........................................................... 18 Table 3.1-5 Atmospheric Dispersion Factors (/Q) (s/m3) - Hope Creek All Accidents .............. 19 Table 3.1-6 Atmospheric Dispersion Factors (/Q) (s/m3) - Salem SGTR Unit 1........................ 19 Table 3.1-7 Atmospheric Dispersion Factors (/Q) (s/m3) All Accidents Except Salem SGTR Unit 1........................................................................................................................................... 20 Table 3.2-1 RG 1.183 Source Term Input Data .......................................................................... 22 Table 3.2-2 RG 1.183 Release Phases ...................................................................................... 22 Table 3.2-3 Core Inventory1 ........................................................................................................ 23 Table 3.2-4 Data for LOCA Model .............................................................................................. 27 Table 3.2-5 Dose Summary for Salem LOCA ............................................................................. 29 Table 3.3-1 Basic Data and Assumptions for SGTR ................................................................... 36 Table 3.3-2 Unit 1 SGTR Accident - Preaccident Iodine Spike Case ........................................ 38 Table 3.3-3 Unit 1 SGTR Accident - Concurrent Iodine Spike Case .......................................... 38 Table 3.3-4 Unit 2 SGTR Accident - Preaccident Iodine Spike Case ........................................ 39 Table 3.3-5 Unit 2 SGTR Accident - Concurrent Iodine Spike Case .......................................... 39 Table 3.4-1 Other Affected Salem Analyses ............................................................................... 40 Table 3.5-1 RG 1.183 Source Term Input Data .......................................................................... 42 Table 3.5-2 RG 1.183 Release Phases ...................................................................................... 42 Table 3.5-3 Core Inventory1 ........................................................................................................ 43 Table 3.5-4 Data for LOCA Model .............................................................................................. 48 Table 3.5-5 EAB Dose Summary for a Hope Creek LOCA ......................................................... 51 Table 3.6-1 Basic Data and Assumptions for MSLB ................................................................... 54 Table 3.6-2 Dose Summary for the MSLB Accident ................................................................... 55 Table 3.7-1 Other Affected Hope Creek Analyses ...................................................................... 56 List of Figures Figure 2.2 Current Site Plan and Exclusion Area Boundary .................................................... 2 Figure 2.6 Proposed Salem Exclusion Area Boundary ........................................................... 5 Figure 2.6 Proposed Hope Creek Exclusion Area Boundary .................................................. 6 Figure 3.1-1 Distances Used in /Q Development For Accident Dose Calculations Hope Creek All Accidents ............................................................................................................................... 15 Figure 3.1-2 Distances Used in /Q Development For Accident Dose Calculations Salem Unit 1 Steam Generator Tube Rupture (SGTR) Accident ..................................................................... 16 Figure 3.1-3 Distances Used in /Q Development For Accident Dose Calculations All Salem Unit 1 and 2 Accidents Except Unit 1 SGTR ...................................................................................... 17

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LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 1

SUMMARY

DESCRIPTION The proposed change updates the dose consequence analyses for Salem Generating Station (Salem) and Hope Creek Generating Station (Hope Creek) to account for changes to the exclusion area to remove those land parcels designated for use by the state of New Jersey for an offshore wind port.

No changes to the Facility Operating Licenses or Technical Specifications are required by this LAR. The Salem and Hope Creek Updated Final Safety Analysis Reports (UFSARs) will be updated in accordance with 10 CFR 50.71(e) as part of implementation of the approved amendment.

2 DETAILED DESCRIPTION Although the updated dose consequences for all analyses remain within the limits of 10 CFR 50.67, the analyses for which the increase in consequences is more than minimal or that result in a departure from a method of evaluation described in the UFSAR require U.S.

Nuclear Regulatory Commission (NRC) approval in accordance with 10 CFR 50.59(c)(2). A summary of the proposed changes is provided below.

2.1 Background

The State of New Jersey is working to develop offshore wind generation capability. The PSEG Nuclear site, with its proximity and access to both safe anchorage and the open ocean, is uniquely suitable for the construction and operation of a marshalling port for offshore wind farms. A marshalling port is an area where components can be staged after arriving from oceangoing vessels before being loaded onto installation vessels.

In 2021, PSEG leased certain parcels within the Salem and Hope Creek Exclusion Area to the New Jersey Economic Development Authority (NJEDA) for construction of the New Jersey Wind Port (NJWP) and for potential future expansion and development of infrastructure to support the offshore wind industry. Salem and Hope Creek have a common exclusion area on land which is coextensive with the Salem and Hope Creek site boundary.

PSEG retains the authority required by 10 CFR Part 100 to determine all activities, including exclusion or removal of personnel and property, within the exclusion area surrounding the reactors, which currently includes the parcels leased to NJEDA.

Both Salem and Hope Creek fully implemented an alternative source term (AST) design bases in accordance with the requirements of 10 CFR 50.67.

2.2 Current Salem and Hope Creek Exclusion Areas The common Salem and Hope Creek land Exclusion Area consists of that area bounded by the site boundary as shown on Figure 2.2-1. For Salem, the minimum distance between the reactors and the exclusion area boundary (EAB) is 1270 meters. For Hope Creek, the minimum distance from the reactor to the EAB is 901 meters.

The exclusion area for each plant extends into the Delaware River. For Salem, the water portion of the exclusion area is defined as the area in the Delaware River within 1270 meters of either the Unit 1 or Unit 2 Containment Building. For Hope Creek, the exclusion area in the Delaware River is the area within 901 meters of the Reactor Building.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 SITE ANJ EXCLUSION NU 80UM)N!Y o 300' &00' 900' t'OO' SC'i£ l)()m 20l)yi 300m 400m SC'i£ PSEC NUCLEAR, LLC SALEM I< HOPE CREEK NUCLEAR GENERATINC STATION CURRENT SITE PLAN I< EXCLUSION AREA BOUNDARY PSEG LAR Figure 2.2 Current Site Plan and Exclusion Area Boundary LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 2.3 Salem Current Design Basis Radiological Analyses Summary Full scope implementation of an AST in accordance with Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Reference 1), was approved in Amendment Nos. 271 for Salem Unit 1 and 252 for Salem Unit 2 (Reference 2).

The current design basis radiological analyses described in the Salem UFSAR are shown below in Table 2.3-1.

Table 2.3-1 Current Salem Design Basis Radiological Analyses Event UFSAR Section Previously docketed Loss of Coolant Accident (LOCA) 15.4.1 ML042740509 Main Steam Line Break (MSLB) 15.4.2 ML042740517 Steam Generator Tube Rupture Accident (SGTR) 15.4.4 ML042740514 Locked Rotor Accident (LRA) 15.4.5 ML042740520 Fuel Handling Accident (FHA) 15.4.6 ML073470363 Rod Ejection Accident (REA) 15.4.7 ML042740511 All of the analyses were subsequently updated from the previously docketed revisions in accordance with 10 CFR 50.59.

2.4 Hope Creek Current Design Basis Radiological Analyses Summary Full scope implementation of an AST in accordance with RG 1.183 was approved in Amendment 134 for Hope Creek (Reference 3).

The current design basis radiological analyses described in the Hope Creek UFSAR are shown below in Table 2.4-1:

Table 2.4-1 Current Hope Creek Design Basis Radiological Analyses Event UFSAR Section Previously docketed Loss of Coolant Accident (LOCA) 15.6.5 ML102371018 Fuel Handling Accident (FHA) 15.7.4 ML063110188 Control Rod Drop Accident (CRDA) 15.4.9 ML101390315 Main Steam Line Break (MSLB) 15.6.4 ML063110168 Instrument Line Pipe Break (ILPB) 15.6.2 ML063110181 Feedwater Line Break outside Containment (FWLB) 15.6.6 ML063110194 With the exception of the ILPB and FWLB, all of the analyses were subsequently updated from the previously docketed revisions in accordance with 10 CFR 50.59.

2.5 Reason for the Proposed Change The current EAB and owner-controlled area (OCA) boundaries are coextensive with the Salem and Hope Creek site boundary. Because the NJWP parcels are located within the current OCA, PSEG is required to include port workers within the onsite emergency response plan and LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 provides port workers with basic training on elements of the PSEG Emergency Plan. The proposed modifications to the EAB remove those areas currently within the Exclusion Area in which NJWP activities, which are unrelated to reactor operation, will be routinely conducted.

Upon implementing the modifications to the EAB, PSEG plans to reduce the OCA to exclude the parcels designated for use by the NJWP, thus supporting PSEG in maintaining its focus on the safe operation of Salem and Hope Creek.

2.6 Description of the Proposed Changes 2.6.1 Modified Exclusion Area Boundary The proposed exclusion areas for Salem and Hope Creek are shown in Figure 2.6-1 and Figure 2.6-2, respectively.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 LEGEND: SITE BOUNDARY

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CIRCULATING WATER INUKE STRUC TURE 0 100m 200m :IOOm 400m SCM.E Figure 2.6 Proposed Hope Creek Exclusion Area Boundary LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 The minimum measured distances to the proposed EAB and the distances used for calculating EAB atmospheric dispersion factors (/Qs) and EAB dose in the updated design basis radiological consequence analyses are shown below:

Table 2.6-1 Minimum Exclusion Area Boundary Distances Minimum Distance EAB Distance Event From to Proposed EAB for Analyses (meters) (meters)

Salem - All RG Center of Salem 1.183 Accidents (excluding 780 695 except S1 SGTR Turbine Building)

S1 Inner 790 790 Penetration Area S1 SGTR S1 Outer 790 790 Penetration Area Hope Creek - All Center of Hope RG 1.183 462 337 Creek Accidents The distances that are used to determine the new EAB /Q values are less than the actual minimum distances to the EAB. These distances are reduced by 10 meters to account for any potential uncertainties.

For Salem events, with the exception of the Salem Unit 1 SGTR event, the minimum EAB distance used in the updated design basis radiological consequence analyses is taken as the distance from the perimeter of a circular accident release boundary with a radius of 85 meters, encompassing all accident release points for Units 1 and 2. For the Salem Unit 1 SGTR event, the minimum distances are taken from the main steam safety valves (MSSVs) located in the Inner and Outer Penetration Areas.

For Hope Creek, the minimum EAB distance used in the updated design basis radiological consequence analyses for all events is conservatively taken as the distance from the perimeter of a circular accident release boundary with a radius of 125 meters, encompassing all accident release points for Hope Creek.

For Salem, the portion of the EAB that extends over the Delaware River is the area included within an arc with a radius 875 meters, encompassing the three circular zones that define the minimum distances to the EAB described above. For Hope Creek, the portion of the EAB that extends over the river is the area within an arc with a radius of 462 meters.

2.6.2 Changes to Design Basis Radiological Consequence Analyses Revised EAB /Qs were developed using meteorological data for the period from 2019 through 2021 for the events and minimum EAB distances listed in Table 2.6-1.

Salem The analyses for the Salem events listed in Table 2.3-1 were revised using the revised /Qs.

The dose consequences for all of the revised analyses remain within the limits of 10 CFR 50.67.

However the revised LOCA analysis results in a more than minimal increase in EAB dose, LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 defined as greater than 10% of the difference between the current calculated dose value and the limits established in 10 CFR 50.67.

The revised SGTR analysis also includes a calculated flashing fraction for the primary coolant (rupture flow) entering the faulted steam generator, and increases in the iodine partition coefficient for both the faulted and intact steam generators.

The MSLB, LRA, FHA and REA events were also affected by the change in EAB /Q but can be implemented without prior NRC approval in accordance with the provisions of 10 CFR 50.59.

Hope Creek The analyses for the Hope Creek events listed in Table 2.4-1 were revised using the revised

/Qs. The dose consequences for all of the revised analyses remain within the limits of 10 CFR 50.67. However the revised LOCA and MSLB analyses result in more than minimal increases in EAB dose.

The revised analyses for the FHA, CRDA, ILPB and FWLB events will be implemented in accordance with the provisions of 10 CFR 50.59.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3 TECHNICAL EVALUATION The Total Effective Dose Equivalent (TEDE) at the EAB for the worst-case two hour period was recalculated for the for the Salem and Hope Creek events in Table 2.3-1 and Table 2.4-1 respectively to account for the revised EAB distances. For the Salem SGTR event, the dose results for the Control Room and Low Population Zone were also recalculated. The calculated radiological consequences were compared with the limits provided in 10 CFR 50.67(b)(2), as clarified in the additional guidance in RG 1.183 for events with a higher probability of occurrence.

3.1 Determination of Atmospheric Dispersion Factors (/Q)

Revised EAB /Qs were developed using the PAVAN computer code. The PAVAN code, documented in NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Plants" (Reference 4) uses the methodology described in RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" (Reference 5).

A comprehensive evaluation of EAB /Q values applicable to the radiological events listed in Tables 2.3-1 and 2.4-1 has been performed for the updated most recent meteorological data and new EAB distances.

Meteorological data is obtained and processed using existing station onsite collection methods and guidance in NRC RG 1.23,"Meteorological Monitoring Programs for Nuclear Power Plants" (Reference 6).

Distances to the nearest point on the updated EAB and the associated accident release points for Salem and Hope Creek were determined. The 95th percentile EAB /Q values were calculated to support accident analyses for Salem and Hope Creek using the PAVAN computer code for the updated distances.

Note that the Control Room and Low Population Zone (LPZ) /Qs are not revised for this activity, as the distances to the Control Rooms and LPZ are not impacted.

Attachments 1 through 5 to this submittal include computer file that contain the site metrological data collected over the years 2019-2021 (Section 3.1.1) and PAVAN input files used in /Q calculations (Section 3.1.2) 3.1.1 Meteorological Data Joint frequency distributions (JFDs) of wind speed, wind direction and Pasquill stability class were developed using the most recent three full years (2019 - 2021) of available hourly meteorological data from the onsite Salem and Hope Creek meteorological tower. The primary tower is a 91 m (300 ft.), guyed lattice structure located at 39° 27 48.9 N, 75° 31 11.76 W.

The NRC previously reviewed PSEG's Onsite Meteorological Measurements Program as part of an Early Site Permit (ESP) Application for the PSEG site, concluding that the onsite meteorological monitoring system provides adequate data to represent onsite meteorological conditions and that onsite data provide an acceptable basis for making estimates of atmospheric dispersion for design-basis accident releases (Reference 7).

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 The Salem and Hope Creek Meteorological Data Acquisition System is designed to meet the intent of RG 1.23, Rev. 1. The meteorological data is compiled and archived hourly as described in RG 1.23. Each year of data provided the hourly wind speed, direction, and sigma theta at 10 m (33 ft.), 46 m (150 ft.), 60 m (197 ft.), and 91 m (300 ft.) as well as wind speed, direction, and sigma theta from the backup meteorological tower at 10 m (33 ft.). Hourly temperature data at 10 m (33 ft.), and 91 m (300 ft.) and dew point temperature at 10 m (33 ft.)

were also obtained with the yearly data. Additionally, hourly T from 10 m - 46 m (33 ft. - 150 ft.), 10 m - 60 m (33 ft. - 197 ft.), and 10 m - 91 m (33 ft. - 300 ft.) was included along with hourly relative humidity values at 10 m (33 ft.) and 91 m (300 ft.). Each year of data was provided in an Excel file providing the date, hour, and meteorological inputs described above, separated into columns.

The JFDs were produced using wind speed and direction measured at 10 m (33 ft.) and the Pasquill stability class based on 10 m - 46 m (33 ft. - 150 ft.) T. Atmospheric stability is distributed into seven stability classes (A - G).

This data was reviewed by meteorologists for missing or anomalous observations. Based on the review of data, no data substitutions were performed. The joint data recovery rate in 2021 was lower than the recovery rates in 2020 and 2019 due to tower maintenance that occurred in November 2021. However, the annual joint data recovery rates exceed the requirement for 90% joint data availability (RG 1.23) for each year and all three years combined. Tower maintenance in 2021 included instrument replacement, in accordance with RG 1.23 criteria, due to obsolescence of existing equipment.

Joint Frequency Distributions The JFDs were developed using the most recent three full years of available meteorological data from the on-site Salem and Hope Creek meteorological tower (2019 - 2021). For consistency with the JFDs developed to support the ESP Application, the updated JFDs were constructed using hourly wind speed and direction recorded at the 10 meter (33 ft.) level and T between 46 meters and 10 meters (150 ft. and 33 ft.).

The annual joint data recovery rates for wind speed/direction and Pasquill stability class exceed the RG 1.23 requirement for 90% joint data recovery for each year.

For the JFDs, wind direction was distributed into 16 directional sectors (i.e., N, NNE, NE, ENE, E, ESE, SE, SSE, S, SSW, SW, WSW, W, WNW, NW, and NNW). Wind speeds were grouped into nominal 0.5 meter per second (m/s) steps from 0.0 m/s to 10.0 m/s and above 10.0 m/s.

The occurrences for the data collected based on directional sector and speed group were then determined. Table 3.1-1 and Table 3.1-2 below provide the distribution of occurrences by wind speed and direction and the frequency of occurrence including all stability classes (A-G).

The wind roses for the most recent data (2019 - 2021) and previous data sets for 2006 - 2008, and 1977 - 2008 are similar. The wind roses show bimodal wind direction distributions with winds from the northwest and southeast occurring most frequently.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.1-1 Pasquill Stability Class A - G Distribution of Occurrences by Speed and Direction SPEED (M/S) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL CALM 0 0.00<WS< 0.50 0 2 1 2 2 0 2 0 1 0 0 1 3 0 3 1 18 0.50<WS< 1.05 41 46 51 63 45 35 24 22 25 30 30 26 32 37 36 39 582 1.05<WS< 1.55 102 110 139 166 155 70 48 70 78 90 71 70 73 98 89 92 1521 1.55<WS< 2.05 154 161 232 202 195 124 100 132 119 156 164 141 131 151 166 137 2465 2.05<WS< 3.05 481 480 509 311 256 315 379 340 329 437 581 477 330 330 539 458 6552 3.05<WS< 4.05 354 336 291 138 99 162 466 321 262 320 389 458 291 282 479 390 5038 4.05<WS< 5.05 251 201 143 70 38 65 493 299 256 208 164 202 238 218 433 298 3577 5.05<WS< 6.05 121 130 67 30 11 21 406 234 179 120 70 87 128 151 379 255 2389 6.05<WS< 8.05 81 69 29 23 7 21 379 193 159 66 42 43 167 224 450 252 2205 8.05<WS< 10.00 12 3 3 0 0 3 61 26 37 16 6 5 52 93 145 52 514 WS> 10.00 0 0 0 0 1 0 43 10 15 3 0 0 18 22 10 1 123 TOTALS 1597 1538 1465 1005 809 816 2401 1647 1460 1446 1517 1510 1463 1606 2729 1975 24984 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.1-2 Frequency of Occurrence

(% of Total Observations) (Calms Distributed In Lowest WS Class)

SPEED (M/S) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.00<WS< 0.50 0.00 0.01 0.00 0.01 0.01 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.01 0.00 0.07 0.50<WS< 1.05 0.16 0.18 0.20 0.25 0.18 0.14 0.10 0.09 0.10 0.12 0.12 0.10 0.13 0.15 0.14 0.16 2.33 1.05<WS< 1.55 0.41 0.44 0.56 0.66 0.62 0.28 0.19 0.28 0.31 0.36 0.28 0.28 0.29 0.39 0.36 0.37 6.09 1.55<WS< 2.05 0.62 0.64 0.93 0.81 0.78 0.50 0.40 0.53 0.48 0.62 0.66 0.56 0.52 0.60 0.66 0.55 9.87 2.05<WS< 3.05 1.93 1.92 2.04 1.24 1.02 1.26 1.52 1.36 1.32 1.75 2.33 1.91 1.32 1.32 2.16 1.83 26.22 3.05<WS< 4.05 1.42 1.34 1.16 0.55 0.40 0.65 1.87 1.28 1.05 1.28 1.56 1.83 1.16 1.13 1.92 1.56 20.16 4.05<WS< 5.05 1.00 0.80 0.57 0.28 0.15 0.26 1.97 1.20 1.02 0.83 0.66 0.81 0.95 0.87 1.73 1.19 14.32 5.05<WS< 6.05 0.48 0.52 0.27 0.12 0.04 0.08 1.63 0.94 0.72 0.48 0.28 0.35 0.51 0.60 1.52 1.02 9.56 6.05<WS< 8.05 0.32 0.28 0.12 0.09 0.03 0.08 1.52 0.77 0.64 0.26 0.17 0.17 0.67 0.90 1.80 1.01 8.83 8.05<WS< 10.00 0.05 0.01 0.01 0.00 0.00 0.01 0.24 0.10 0.15 0.06 0.02 0.02 0.21 0.37 0.58 0.21 2.06 WS> 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.17 0.04 0.06 0.01 0.00 0.00 0.07 0.09 0.04 0.00 0.49 TOTALS 6.39 6.16 5.86 4.02 3.24 3.27 9.61 6.59 5.84 5.79 6.07 6.04 5.86 6.43 10.92 7.91 100.00 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Pasquill Stability Class.

Temperature measurements determine the Pasquill stability class at the site. The Pasquill stability class is an effective indicator of worst-case stability conditions (Pasquill class E, F and G).

The frequency of occurrence of the respective Pasquill stability class distributions is consistent among the three time periods evaluated. Stability class E occurred most frequently, followed by stability class D. Stability classes B and C occurred least frequently. The 2019 - 2021 data are consistent with the previous evaluations cited in the Site Safety Analysis Report for the PSEG Site ESP application (Reference 8).

3.1.2 EAB /Q The PAVAN code (Reference 4) is used to calculate the 95th percentile short-term relative air concentration (/Q) factors for Hope Creek and Salem to support the updated accident analyses. PAVAN determined /Q values at the new EAB minimum distance for each of the sixteen downwind direction sectors and the overall site to find the maximum 0-2 hour /Q.

A new EAB /Q for Salem was determined and the value used in the accident analyses is 2.44E-4 s/m3 (a value of 1.97E-04 s/m3 is applicable for a Unit 1 Steam Generator Tube Rupture accident). A new EAB /Q for Hope Creek was determined and the value used in the accident analyses is 8.14E-4 s/m3.

Offsite EAB Distance - Hope Creek Generating Station The minimum distance at which Hope Creek EAB doses are calculated is 337 m from a 125 m circle that encompasses the Hope Creek Reactor, Auxiliary/Control, and Turbine Buildings (462 m from the center point).

The 125 m circle encompasses all release points from all design basis accidents analyzed. It is noted that the circle extends beyond the release point for some accidents and therefore can be considered conservative. The closest point from the 125 m release circle to the nearest point on the new EAB is 337 m. Other edges of the EAB are greater than 337 m.

Offsite EAB Distance - Salem Generating Station The minimum distance to the new EAB from an 85 m radius circle encompassing all Salem Unit 1 and 2 design basis accident release points (with the exception of the Unit 1 SGTR) is 695 m.

The 85 m radius circle encompasses all release points from all design basis accidents analyzed with the exception of SGTR for Salem Unit 1. It is noted that the circle extends beyond the release point for some accidents and therefore can be considered conservative. The closest point from the 85 m release circle to the nearest point on the new EAB is 695 m. Other edges of the EAB are greater than 695 m.

The minimum distance from the limiting Salem Unit 1 SGTR accident release point (Inboard MSSVs) to the new EAB on land is 790 m. Other edges of the EAB are greater than 790 m.

Over water, the Salem EAB is a circular zone of radius 875 m. The center of the area containing the Salem Reactor Building, Auxiliary Building, Fuel Handling Building and Tanks is also the center of the circular zone. Note that this circular zone also encompasses the 695 m LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 and two 790 m circles with respect to the locations of the inboard and outboard MSSVs. Thus, Salem /Q values are calculated at 790 m for the Unit 1 SGTR accident, and at 695 m for the remaining Salem Design Basis Accidents.

Figure 3.1-1 for Hope Creek and Figure 3.1-2 and Figure 3.1-3 for Salem are provided to show the distances described above.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Figure 3.1-1 Distances Used in /Q Development for Accident Dose Calculations Hope Creek All Accidents LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Figure 3.1-2 Distances Used in /Q Development for Accident Dose Calculations Salem Unit 1 Steam Generator Tube Rupture (SGTR) Accident LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Figure 3.1-3 Distances Used in /Q Development for Accident Dose Calculations All Salem Unit 1 and 2 Accidents except Unit 1 SGTR LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Summary of PAVAN Inputs and /Q Results Table 3.1-3 and Table 3.1-4 include PAVAN input parameters related to instrument elevations, release heights, and Containment parameters used to determine building wake effects for Hope Creek and Salem, respectively. Building wake effects were considered in the existing methodology for both stations.

Table 3.1-3 Hope Creek Summary of Common PAVAN Inputs Description Value No terrain adjustment factors are applied. Site is considered to be flat for this analysis. 1.0 The minimum (vertical-plane) cross-sectional area (square meters) of the structure associated with the release point. 2940 The height (meters) above plant grade of the structure used in the building-wake term for the annual average calculations. 61.7 The height (meters) above the plant grade of the release point. Since this analysis is for a ground release, a value of 10 m is used. 10 The height (meters) above ground-level at which the wind speed was measured. 10 Table 3.1-4 Salem Summary of Common PAVAN Inputs Description Value No terrain adjustment factors are applied. Site is considered to be flat for this analysis. 1.0 The minimum (vertical plane) cross-sectional area (square meters) of the structure associated with the release point. 2399 The height (meters) above plant grade of the structure used in the building-wake term for the annual average calculations. 58.3 The height (meters) above the plant grade of the release point. Since this analysis is for a ground release, a value of 10 m is used. 10.0 The height (meters) above ground-level at which the wind speed was measured. 10.0 Table 3.1-5 for Hope Creek and Table 3.1-6 and Table 3.1-7 for Salem provide the 95th percentile EAB /Qs for the most limiting 0-2 hour period. The most limiting sector value (i.e.,

largest /Q value) in each table is provided in bold text. Site overall /Qs were calculated in accordance with RG 1.145 and are less than the most limiting sector values.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.1-5 Atmospheric Dispersion Factors (/Q) (s/m3) - Hope Creek All Accidents Downwind Sector Distance (m) 0-2 hours S 337 7.73E-04 SSW 337 8.14E-04 SW 337 7.44E-04 WSW 337 6.69E-04 W 337 6.55E-04 WNW 337 6.74E-04 NW 337 7.90E-04 NNW 337 5.65E-04 N 337 4.39E-04 NNE 337 4.62E-04 NE 337 5.46E-04 ENE 337 4.53E-04 E 337 4.34E-04 ESE 337 4.18E-04 SE 337 6.11E-04 SSE 337 6.37E-04 Table 3.1-6 Atmospheric Dispersion Factors (/Q) (s/m3) - Salem SGTR Unit 1 Downwind Sector Distance (m) 0-2 hours S 790 1.88E-04 SSW 790 1.97E-04 SW 790 1.85E-04 WSW 790 1.69E-04 W 790 1.65E-04 WNW 790 1.74E-04 NW 790 1.86E-04 NNW 790 1.48E-04 N 790 1.21E-04 NNE 790 1.24E-04 NE 790 1.46E-04 ENE 790 1.29E-04 E 790 1.20E-04 ESE 790 1.20E-04 SE 790 1.64E-04 SSE 790 1.67E-04 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.1-7 Atmospheric Dispersion Factors (/Q) (s/m3)

All Accidents except Salem SGTR Unit 1 Downwind Distance Sector (m) 0-2 hours S 695 2.32E-04 SSW 695 2.44E-04 SW 695 2.29E-04 WSW 695 2.08E-04 W 695 2.04E-04 WNW 695 2.14E-04 NW 695 2.31E-04 NNW 695 1.82E-04 N 695 1.50E-04 NNE 695 1.54E-04 NE 695 1.80E-04 ENE 695 1.60E-04 E 695 1.47E-04 ESE 695 1.47E-04 SE 695 2.02E-04 SSE 695 2.06E-04 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.2 Salem Loss of Coolant Accident (LOCA)

This section describes the methods employed and results obtained from the radiological reanalysis of the design basis LOCA for Salem. The analysis considers dose contributions from the following post-LOCA release paths:

Containment Leakage

Engineered Safety Feature (ESF) Leakage

Containment Vacuum Relief Line Release

Back-leakage to the Refueling Water Storage Tank (RWST) Leakage Doses are calculated at the EAB for the worst-case two-hour period. The methodology used to evaluate the EAB doses resulting from a LOCA is consistent with RG 1.183 (Reference 1).

3.2.1 LOCA 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 RG 1.183, 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, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 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 from there to the environment via paths listed in the section above.

3.2.2 LOCA Source Term Definition RG 1.183 provides explicit description of the key AST characteristics recommended for use in design basis radiological analyses. The core radionuclide inventory used in this analysis corresponds to 3,632 MWt, which is 105% of the current licensed thermal power of 3,459 MWt.

Table 3.2-1 lists the RG 1.183 source term inputs used in the LOCA analysis and includes: the core inventory release fractions by radionuclide group, timing of release, and chemical form of the release into Containment.

RG 1.183 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 in the subsequent 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Table 3.2-2 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 3.2-3 lists the isotopes and the associated power-normalized curies at the end of a fuel cycle that were input into RADTRAD. The Committed Effective Dose Equivalent (CEDE) and Effective Dose Equivalent (EDE) dose conversion factors used for each of the isotopes were based on Federal Guidance Reports (FGRs) 11 and 12 (References 9 and 10 respectively).

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.2-1 RG 1.183 Source Term Input Data Characteristic RG 1.183 Source Term Noble Gases 100%

Iodine 40%

Core Fractions Released To Cesium 30%

Containment Tellurium 5%

Barium 2%

Others -0.02% to 0.25%

Released in Two Phases Over Timing of Release 1.8-hour Interval 4.85% Elemental Iodine Chemical and Physical 0.15% Organic Vapor Form 95% Aerosol Solids Treated as an Aerosol Table 3.2-2 RG 1.183 Release Phases Core Release Fractions1 Early In-Isotope Group Gap Vessel Noble Gases2 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium 0 0.05 Barium, Strontium 0 0.02 Noble Metals 0 0.0025 Cerium 0 0.0005 Lanthanides 0 0.0002 Duration (hours) 0.5 1.3 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.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.2-3 Core Inventory1 Isotope2 Ci/MWt Isotope2 Ci/MWt Isotope2 Ci/MWt Co-58 2.553E+02 Ru-103 3.958E+04 Cs-136 1.146E+03 Co-60 1.953E+02 Ru-105 2.574E+04 Cs-137 2.107E+03 Kr-85 3.056E+02 Ru-106 8.993E+03 Ba-139 5.474E+04 Kr-85m 7.222E+03 Rh-105 1.783E+04 Ba-140 5.416E+04 Kr-87 1.306E+04 Sb-127 2.429E+03 La-140 5.535E+04 Kr-88 1.861E+04 Sb-129 8.602E+03 La-141 5.077E+04 Rb-86 1.646E+01 Te-127 2.345E+03 La-142 4.894E+04 Sr-89 3.128E+04 Te-127m 3.105E+02 Ce-141 4.924E+04 Sr-90 1.689E+03 Te-129 8.075E+03 Ce-143 4.787E+04 Sr-91 4.022E+04 Te-129m 2.129E+03 Ce-144 2.967E+04 Sr-92 4.186E+04 Te-131m 4.078E+03 Pr-143 4.700E+04 Y-90 1.812E+03 Te-132 4.059E+04 Nd-147 2.102E+04 Y-91 3.812E+04 I-131 2.750E+04 Np-239 5.632E+05 Y-92 4.201E+04 I-132 3.889E+04 Pu-238 3.192E+01 Y-93 4.752E+04 I-133 5.556E+04 Pu-239 7.200E+00 Zr-95 4.815E+04 I-134 6.111E+04 Pu-240 9.079E+00 Zr-97 5.018E+04 I-135 5.278E+04 Pu-241 1.529E+03 Nb-95 4.552E+04 Xe-133 5.556E+04 Am-241 1.010E+00 Mo-99 5.313E+04 Xe-135 1.389E+04 Cm-242 3.865E+02 Tc-99m 4.586E+04 Cs-134 3.768E+03 Cm-244 2.262E+01 Notes:

1. Power level = 3,632 MWt (105% of 3,459 MWt, rated thermal power)

2. Noble Gasses (Xe, Kr), Halogens (I), Alkali Metals (Cs, Rb), Tellurium Group (Te, Sb, Ba, Sr),

Noble Metals (Ru, Rh, Mo, Tc, Co), Lanthanides (La, Zr, Nd, Nb, Pr, Y, Cm, Am), Cerium (Ce, Pu, Np) 3.2.3 LOCA Atmospheric Dispersion Factors LOCA Offsite/EAB /Q The EAB atmospheric dispersion factor is 2.44E-04 s/m3.

3.2.4 Containment Leakage Activity Transport in Primary Containment The Salem containment spray covers 75% of the containment volume. Five containment fan cooling units (CFCUs) are to be operable per TS 3.6.2.3. The CFCUs take suction from the area above the operating floor and discharge into a ring header located just below the operating floor. The ventilation ring header distributes air to both the sprayed and unsprayed regions in the containment. During normal operation, the 5 CFCUs deliver 550,000 cfm to the various containment areas. Of the 550,000 cfm of normal flow, approximately 185,000 cfm is distributed above the operating floor and 60,000 cfm is distributed to the steam generator cubicles.

Therefore, a total of 245,000 cfm out of the 550,000 cfm of CFCUs air is distributed to sprayed regions, which is 44.55% of the total CFCUs flow (245,000/550,000 x 100% = 44.55%).

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 During an accident each CFCU is designed to supply a nominal flow of 39,000 cfm. Only two CFCUs are conservatively assumed to be available during a LOCA, which provide a total air circulation flow of 78,000 cfm (2 x 39,000 cfm = 78,000 cfm). Assuming that the post-LOCA distribution of air to the various areas remains the same as that in the normal operating condition, the operation of two CFCUs will distribute 55.45% of the total air circulation flow of 78,000 cfm, which is 43,251 cfm, from the sprayed region to the unsprayed regions. The air circulation between both the sprayed and unsprayed regions is accounted for by also using the circulation flow of 43,251 cfm from the unsprayed region to the sprayed region.

Per RG 1.183, Appendix A, Section 3.3, the mixing rate between sprayed and unsprayed regions of the containment building is assumed to be two turnovers of the unsprayed region volume due to natural convection, unless another rate is justified. The conservatively low post-LOCA two CFCU air flow of 43,251 cfm yields approximately four turnovers of the unsprayed region per hour [4.0 = (43,251 ft3/min x 60 min/hr) / (0.25 x 2.62E+06 ft3)].

Extended Containment Spray Operation Containment spray removal of iodine and particulates (aerosols) is assumed to be initiated at 90 seconds after the start of the LOCA event. In the initial containment spray injection phase the spray water is drawn from the refueling water storage tank (RWST). The injection phase terminates at 4141.6 seconds, or 69 minutes but is assumed to terminate at 48 minutes after the start of the LOCA, at which time the containment spray recirculation phase begins and the spray water is drawn from the containment sump. It is conservative to use the shorter injection spray duration of 48 minutes since the spray removal rates are larger during the injection phase than during the recirculation phase. Using the shorter duration for the injection phase means there is more airborne activity in containment atmosphere available for release which maximizes activity releases from containment leakage which translates into higher off-site and control room doses.

A containment spray interruption of 10 minutes is assumed during the transition from the injection to the recirculation phase (from 48 to 58 minutes). During the interruption, the containment spray is not credited in removing the aerosol and elemental iodine activities from the containment atmosphere. The containment spray injection phase elemental removal coefficient is calculated to be 29 hr-1 using the Salem plant-specific design condition parameters. However, Standard Review Plan (SRP) 6.5.2 (Reference 11) limits the use of the elemental iodine removal coefficient E to 20 hr-1. Therefore, a E of 20 hr-1 is used during the injection phase. The spray aerosol removal coefficient P in the injection phase is calculated to be 4.44 hr-1 using the plant-specific design input information. During the injection phase, one containment spray pump is assumed to operate (postulating a single failure of the other pump) at a minimum flow rate of 2,600 gpm. The containment spray flow rate is reduced to 1,900 gpm during the recirculation phase. Therefore, for the recirculation phase E and P are re-calculated based on the reduced flow rate.

SRP 6.5.2 sets forth a maximum decontamination factor (DF) of 200 for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. RG 1.183 specifies that the maximum activity to be used in determining the containment spray DF is defined as the iodine activity in the columns labeled "Total" in Table 2 of RG 1.183 multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosols are treated as particulates in SRP methodology). The DF for the containment atmosphere achieved by the containment spray system is calculated to be 546 using the following equation:

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 VS H

= 1+

DF VC Where: Vs = volume of liquid in containment sump Vc = containment net free volume less Vs H = partition coefficient A DF for spray cutoff time for elemental iodine of 100 (which is smaller than both 200 and 546) is conservatively used in the containment leakage analysis. The SRP 6.5.2 also states that the particulate iodine removal rate should be reduced by a factor of 10 when a particulate DF of 50 is reached.

The maximum iodine activity in the containment atmosphere occurs at the end of early-in-vessel release phase, which is 1.8 hrs after a LOCA. Therefore, the corresponding E and P cutoff times are calculated based on an assumption that the maximum iodine activity occurs at 1.8 hrs after onset of a LOCA. Although one containment spray can operate for a long time after a LOCA, containment spray operation is assumed to be terminated at 4.0 hrs after a LOCA.

Long-Term Iodine Partition RG 1.183, Appendix A, Section 1, requires evaluation of the re-evolution of iodine for a sump pH value of less than 7. During the injection and recirculation phases, the sump water pH will remain at > 7 including the effect of acids and bases created during the LOCA event and radiolysis products. Consequently, the re-evolution of dissolved iodine from the sump is not considered.

3.2.5 Engineered Safety Feature (ESF) System Leakage ESF systems that recirculate containment sump water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components in the Residual Heat Removal (RHR), Safety Injection (SI), Containment Spray (CS) and Chemical and Volume Control (CVC) systems. The RHR pumps provide the suction boost and sump water supply to the SI pumps, centrifugal charging pumps (CCPs), and containment spray headers during the recirculation phase. The Positive Displacement Pump (PDP), which is a potential major source of leakage during normal operating conditions, is not required to be operational during a LOCA. EOP-LOCA-3 directs stopping the PDP if running prior to opening the SJ45 valves, which feed containment sump water from the discharge of the RHR pumps downstream of the RHR heat exchangers to the suction headers of the SI and CVC pumps and to a containment spray header during the recirculation phase. The PDP is stopped, but it is not isolated from the recirculation flow path. Although, the PDP pump does not run during the post-LOCA recirculation phase, its leak rate contribution is analyzed as a part of the ESF leakage.

Doing so provides an analyzed basis for PDP leakage during normal operation so that the PDP pump can be operated at a relatively higher leak rate, which allows for using the PDP as the preferred normal charging pump. Ultimately, the SI pumps and RHR pumps become the most likely sources of the ESF leakage during the post-LOCA recirculation mode.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 The radiological consequences from the postulated leakage are analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.

3.2.6 Containment Vacuum Relief Line Release The containment vacuum relief line release (VRLR) occurs following the Large Break LOCA, and before containment isolation. The entire RCS inventory is assumed to be instantaneously released and homogeneously mixed in the containment atmosphere. The containment pressurization due to the RCS mass and energy release combined with the presence of the containment vacuum relief line result in the potential for a limited release of airborne activity to the environment. 100% of the radionuclide inventory in the RCS liquid is assumed to be released into the containment at the initiation of the LOCA. A release of gap activity into the containment is not considered since the containment vacuum relief line release duration of 5 seconds terminates this release path prior to the onset of the gap release phase.

The RCS activity inventory is based on the technical specification for RCS equilibrium activity.

Iodine spikes are not considered. Per the TS Limiting Condition for Operation (LCO) for Specific Activity, the primary coolant equilibrium iodine concentration permitted by the TS is 1 µCi/gm Dose Equivalent (DE) I-131. TS Section 1.10 defines DE I-131 as follows:

DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (µCi/gm) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present.

This analysis calculates DE I-131 in terms of the thyroid dose conversion factors specified in FGR 11 in accordance with the TS Section 1.10 definition.

A 1% failed fuel primary coolant iodine concentration yields 3.58 µCi/gm DE I-131, which is conservatively greater than the TS limit of 1.0 µCi/gm DE I-131. The RCS noble gas concentrations corresponding to 1% fuel defects are conservatively used to determine the total RCS noble gas activity released in the containment during the VRLR based on the total coolant mass in the RCS. The isotopic activity is divided by the core thermal power level of 3,632 MWt to obtain the isotopic Ci/MWt for input into RADTRAD.

3.2.7 Post-LOCA Back-leakage to the Refueling Water Storage Tank (RWST)

ESF systems that recirculate containment sump water outside of the primary containment are assumed to back-leak to the RWST during their intended operation. The iodine release from the ESF back-leakage to the RWST is released via the tank vent to the environment without any credit taken for plateout of elemental iodine on the tank surface.

ESF system back-leakage to the RWST is conservatively assumed to begin at 0 minutes with the ECCS operating at maximum capacity. The activity release to the environment from the RWST vent is also conservatively assumed to begin at 0 minutes. A worst-case back-leakage to the RWST is assumed to be 2.9 gpm.

3.2.8 Data for LOCA Model Table 3.2-4 provides a summary of parameters used to reanalyze the EAB LOCA dose analysis at Salem.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.2-4 Data for LOCA Model Parameter Value Used Activity Transport in Primary Containment Containment Net Free Volume 2.62E+06 ft3 Sprayed Volume 75% of containment volume 1.965E+06 ft3 Unsprayed Volume 25% of Containment volume 6.55E+05 ft3 Containment Leak Rate 0.1 v%/day 0-24 Hrs 0.05 v%/day 24-720 Hrs Containment Fan Cooler Unit (CFCU) Flow 39,000 cfm/fan Rate Flow Rate From Sprayed To Unsprayed 43,251 cfm (2 of 5 CFCUs operational)

Volume Flow Rate From Unsprayed To Sprayed 43,251 cfm (2 of 5 CFCUs operational)

Volume Spray Initiation Time 85 sec 90 sec used in the analysis Spray Recirculation Phase Initiation Time 48 min used in containment leakage analysis Maximum Iodine Activity Occurs in 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Containment Containment Spray Injection Phase Flow 2,600 gpm (1 train at 2,600 gpm)

Rates Spray Removal Coefficients In Injection Phase Elemental (E2600) 20 hr-1 Particulate (P2600) (Aerosol) 4.44 hr-1 Containment Spray Flow Rate During 1,900 gpm Recirculation Phase Spray Removal Coefficients In Recirculation Phase Elemental (E1900) 14.62 hr-1 Particulate (P1900) (Aerosol) 3.24 hr-1 Decontamination Factor (DF) For Spray Cutoff Time Elemental 100 Particulate (Aerosol) 50 Spray Interruption During Transition From 5 minutes Injection To Recirculation Phase 10 minutes used in the analysis Post-LOCA Spray Removal Coefficients (hr-1) During Gap and Early-In-Vessel Release Phases for Containment Leakage Analysis Time (hour) Aerosol Elemental 0 0 0 0.025 (90 seconds) 4.44 20 0.800 (48 minutes) 0 0 0.967 (58 minutes) 3.24 14.62 2.115 3.24 0 3.007 0.324 0 4.0 0 0 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Parameter Value Used ESF Leakage Parameters Sump Water Volume 43,930 ft3 @ > 48 minutes ESF Leakage Rate 0.45 gpm (0.9 gpm used in the analysis)

ESF Leakage Initiation Time 0 minutes (with ECCS operating at maximum capacity)

Sump Water pH > 7.0 ESF Leakage Iodine Flashing Factors:

0 to 1.0 hour0 days 0 hours 0 weeks 0 months 0.0526 (5.26%)

1.0 to 16.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months 0.0376 (3.76%)

> 16.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months 0.02 (2 %)

Chemical Form of Iodine In ESF Leakage Elemental 97%

Organic 3%

Fraction of Core Iodine In Sump Water 40%

Auxiliary Building Ventilation System (ABVS) 0.0% (assumed for elemental, organic and Charcoal Filter Efficiencies particulate iodine)

Containment Sump Water Temperature 262.39°F (maximum) @ 1.399E+03 sec 248.11°F (@ 3.599E+03 sec).

< 170°F (@ 5.9999E+04 sec)

Site Boundary Release Model Parameters EAB Atmospheric Dispersion Factor (/Q) 2.44E-04 s/m3 EAB Breathing Rate (m3/sec) 3.5E-04 EAB allowable dose limit 25 rem TEDE for any 2-hour period Containment Vacuum Relief Line Release Parameters Relief Valve Closure Time 2.0 sec (5.0 sec used in analysis)

Volume of Containment Air Release Via 1,014 ft3 Vacuum Relief Line Reactor Coolant System Volume 12,707 ft3 @ a nominal Tavg of 561°F Maximum Coolant (1% Failed Fuel) Reactor Activity Values Isotope µCi/g Isotope µCi/g Isotope µCi/g 4.0E-01 Xe- 1.7E+01 I-131 2.8E+00 Kr-83m 133m Kr-85m 1.7E+00 Xe-133 2.6E+02 I-132 2.8E+00 8.2E+00 Xe- 4.9E-01 I-133 4.2E+00 Kr-85 135m Kr-87 1.0E+00 Xe-135 8.5E+00 I-134 5.7E-01 Kr-88 3.0E+00 Xe-137 1.8E-01 I-135 2.3E+00 Xe-131m 2.1E+00 Xe-138 6.1E-01 RWST Back-leakage Parameters RWST Back-leak Rate I 2.9 gpm RWST Minimum Air Volume I 51,696 ft3 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.2.9 Salem LOCA Results Table 3.2-5 lists the TEDE dose at the EAB from a LOCA at Salem. The dose at the EAB is less than the 25 rem TEDE limit stated in 10 CFR 50.67 and RG 1.183. The EAB dose represents the worst 2-hour dose for each release pathway.

Table 3.2-5 Dose Summary for Salem LOCA Post-LOCA Activity Release Path EAB Post-LOCA TEDE (Rem)

Containment Leakage Sprayed Region 3.0214E+00 Unsprayed Region 1.5319E+00 Total 4.5533E+00 (occurs @ t = 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ) 2.0637E+00 ESF Leakage (occurs @ t = 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) 3.9795E-04 Containment Relief Line Release (occurs @ 0.0 hours0 days 0 hours 0 weeks 0 months ) 1.6913E-02 RWST Back-Leakage (2.9 gpm back-leakage)

(occurs @ t = 96.0 hours0 days 0 hours 0 weeks 0 months ) 2.298E-01 Containment Shine (occurs @ t =1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

Total 6.86+00 Allowable TEDE Limit 2.50E+01 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.3 Salem Steam Generator Tube Rupture (SGTR) Accident This section describes the methods employed and the results of the SGTR design basis radiological analysis. This analysis includes dose consequences associated with the releases of the radioactive material initially present in primary coolant and secondary coolant at maximum allowable TS concentrations plus iodine spiking. Doses are calculated at the EAB for the worst-case two-hour period. The methodology used to evaluate the doses resulting from a SGTR is consistent with RG 1.183.

3.3.1 SGTR Scenario Description A SGTR is a break in a tube carrying primary coolant through the steam generator (SG). This postulated break allows primary coolant to leak to the secondary side of one of the steam generators (denoted as the affected or faulted steam generator) that results in an assumed release to the environment. This event assumes the complete severance of a single SG tube.

With this exception, the RCS pressure boundary remains intact. This assessment addresses the reactor coolant activity concentrations corresponding to (1) a pre-accident iodine spike and (2) a concurrent iodine spike. There is no fuel failure during this event. The SG liquid masses and steam mass releases are different for Salem Units 1 and 2 SGs. Therefore, the SGTR accident is analyzed separately for each unit.

Pre-accident Iodine Spike Release In the pre-accident iodine spike release scenario, a reactor transient has occurred prior to the postulated SGTR accident and has raised the primary coolant (PC) (also referred to as RCS) iodine concentration to the maximum value permitted by technical specifications. The PC iodine concentration of 60 µCi/gm Dose Equivalent (DE) of I-131 is based on TS 3.4.8 for Unit 1 and TS 3.4.9 for Unit 2. The plant-specific PC iodine concentration profile corresponds to 1% fuel defects. The iodine dose conversion factors are used to determine scaling factors for the iodine isotopes which are used to convert the iodine concentration of 1% fuel defects to 1 µCi/gm DE I-131. The total activities for iodine isotopes in Units 1 and 2 PC are conservatively calculated using the PC mass determined at cooled liquid conditions.

The PC noble gas concentrations corresponding to 1% fuel defects are conservatively used to determine the total noble gas activities in Units 1 and 2 PC. The total iodine plus noble gas isotopic activities released to the Units 1 and 2 PC are calculated and then used to develop the Units 1 and 2 RADTRAD Nuclide Inventory Files (NIFs) for the pre-accident iodine spike cases.

The following paragraphs describe the activity transport paths from the one faulted and three intact SGs. These activity transport paths are applicable to both the pre-accident and concurrent iodine spike cases.

Faulted SG Release The PC is postulated to be released into the faulted SG having a single SG tube ruptured. The resultant equilibrium break flow is assumed to continue for 30 minutes after the initiation of the event, at which time it is assumed that the operator isolates the faulted SG and depressurizes the RCS to the faulted SG pressure ending the transfer of PC to the secondary side of the faulted SG and ending the release to the atmosphere. To maximize the offsite doses, it is assumed that offsite power is lost so that the main steam condensers are not available. The LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 iodine activity in the primary-to-secondary (P-T-S) leakage is assumed to leak at a rate of 1 gpm into the faulted SG for 0-0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and then it continues to leak into the remaining three (3) intact SGs for 0.5-32 hrs. Iodine activity in the PC released from the ruptured tube that does not immediately flash and in the P-T-S leakage is diluted in the bulk water of the faulted SG and released from the faulted SG to the environment in proportion to the steaming rate and the iodine partition coefficient value of 100 recommended in RG 1.183, Appendix F, Section 5.6, and Appendix E, Section 5.5.4. The current SGTR design basis radiological analysis uses partition coefficients of 1 for P-T-S rupture flow in the faulted SG and 10 for secondary liquid iodine release in the faulted SG. Iodine activity in the PC released from the ruptured tube that immediately flashes to vapor is assumed to be released without decontamination (i.e., no credit is taken for scrubbing as the vapor rises through the bulk water of the SG). The holdup time for the flashed vapor is negligible as the SG steam space is assumed to be 1 ft3. The current SGTR design basis radiological analyses does not model flashing of PC released from the ruptured tube in the faulted SG.

PC noble gas activity entering the faulted SG is assumed to be released to the environment without holdup or decontamination in the secondary system. For the secondary liquid iodine release from the faulted SG, an iodine partition coefficient of 100 is used consistent with RG 1.183, Appendix F, Section 5.6, and Appendix E, Section 5.5.4, and the SG steaming rates are reduced accordingly for the secondary liquid iodine release.

Intact SGs Release Due to the SGTR transient, the reactor shuts down and the RCS loses a substantial amount of coolant inventory with the potential of over flooding the faulted SG. In order to remove decay heat, the plant begins to release the secondary coolant via the intact SGs' atmospheric relief valves. The steam release from the three intact SGs continues for 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months until the RHR system is aligned to dissipate heat. During the 0.5-32 hours, the PC is assumed to leak into the three intact SGs. During the 0.5-32 hours, the total P-T-S leak rate into the three intact SGs for Salem is conservatively assumed to be one gpm, which is greater than the 0.313 gpm corresponding to three times the maximum TS leak rate of 150 gal/day through any one SG

[0.313 gpm = (3 SGs) x (150 gal/day per SG) / (1440 min/day)].

The PC noble gases are released directly to the environment from the RCS without holdup. The iodine activity introduced into the SGs via the P-T-S leakage is assumed to be diluted within the SGs liquid. This is a valid assumption based on a study documented in WCAP-13132 (Reference 12) that concludes that the overall probability of a significant radioiodine release during a SGTR transient, including the effects of SG tube uncovering, is sufficiently low to exclude such events from consideration. Therefore, the recommended value of 100 in RG 1.183, Appendix F, Section 5.6 and Appendix E, Section 5.5.4, for the iodine partition coefficient is used in this analysis and the SG steaming rates are reduced accordingly. For the secondary liquid iodine release from the three intact SGs, an iodine partition coefficient of 100 is used consistent with RG 1.183 and the SG steaming rates are reduced accordingly. The current SGTR design basis radiological analysis uses a partition coefficient of 10 for secondary liquid iodine release in the intact SGs.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Concurrent Iodine Spike In the concurrent iodine spike release scenario, the primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in PC iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the PC (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in TS. The isotopic iodine appearance rates are calculated using the specific activity production and removal rates based on the most conservative letdown system parameters. The total isotopic iodine activity is calculated using the spike duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The total isotopic iodine and noble gas activities are used to develop the RADTRAD NIFs for the concurrent iodine spike cases for Units 1 and 2. The activity in the fuel is postulated such that 99% of all fuel activity is released into the RCS within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months at a calculated rate of 9.60 cfm. The activity transport from the RCS to the environment is the same as that for the pre-accident iodine spike case.

Steam Generator Liquid Iodine Activity Release The iodine activity in the SG liquid is calculated based on a secondary side liquid activity concentration of 0.1 µCi/gm DE I-131 using the Units 1 and 2 SG liquid masses. The iodine partition coefficient of 100 is used consistent with RG 1.183. The iodine activity is assumed to be released to the environment in proportion to the steaming rate and the iodine partition coefficient value.

3.3.2 SGTR Source Term There is no fuel damage postulated for the SGTR accident. Per RG 1.183, Appendix F, Section 2, since no or minimal fuel damage is postulated for the limiting event, the released activity is the maximum coolant activity allowed by the TS. Two cases of iodine spiking, a pre-accident case and a concurrent case, are assumed.

Consistent with RG 1.183, Appendix F, Section 3, the activity released from the fuel (for both the pre-accident and concurrent iodine spike cases) is assumed to be released instantaneously and homogeneously through the PC. In addition, consistent with RG 1.183, Appendix F, Section 4, the chemical form of iodine releases from the SGs to the environment are assumed to be 97% elemental and 3% organic. These fractions apply to iodine released during normal operations, including iodine spiking.

Pre-Accident Iodine Spike Source Term Consistent with RG 1.183, Appendix F, Section 2.1, the first iodine spiking case assumes that a reactor transient has occurred prior to the postulated SGTR and has raised the PC iodine concentration to the maximum value permitted by the TS.

Per the TS LCO for Specific Activity, the maximum primary coolant iodine concentration permitted (for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after discovering the specific activity is > 1 µCi/gm DE I-131) is 60 µCi/gm DE I-131. TS 1.10 defines DE I 131 as that concentration of I-131 (in µCi/gm) that alone would produce the same dose when inhaled as the combined activities of I-131, I-132, I-133, I-134, and I-135 actually present using the thyroid committed dose equivalent conversion factors specified in FGR 11.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Concurrent Iodine Spike Source Term Consistent with RG 1.183, Appendix F, Section 2.2, the second iodine spiking case assumes that the primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in PC iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the PC (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in TS. The assumed iodine spike duration is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Per the TS LCO for Specific Activity, the equilibrium PC iodine concentration permitted by the TS is 1.0 µCi/gm DE I-131.

Noble Gas Source Term In both pre-accident and concurrent iodine spike cases the PC noble gas (Xenon and Krypton) concentrations are assumed to correspond to the maximum coolant activity allowed by the TS.

Per the TS LCO for Specific Activity the equilibrium PC non-iodine concentration permitted by the TS is 600 µCi/gm DE Xe-133. TS 1.11 defines DE Xe-133 as that concentration of Xe-133 (in µCi/gm) that alone would produce the same acute dose to the whole body as the combined activities of Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present using the effective dose conversion factors for air submersion in FGR 12.

A 1% failed fuel PC iodine concentration is 3.58 µCi/gm DE I-131, which is conservatively greater than the TS limit of 1.0 µCi/gm DE I-131. Since 1% failed fuel introduces more iodine activity into the PC than is allowed by the TS, it can be conservatively assumed that 1% failed fuel introduces more non-iodine activity into the RCS than is allowed by the TS limit for DE Xe-133. Therefore, the PC noble gas concentrations corresponding to 1% fuel defects are conservatively used to determine the total RCS noble gas activity.

3.3.3 SGTR Release Transport Primary-to-Secondary (P-T-S) Leak Rates Consistent with RG 1.183, Appendix F, Section 5.1, the P-T-S leak rate is apportioned between affected and unaffected SGs in such a manner that the calculated dose is maximized. The TS LCO for RCS Operational Leakage specifies a maximum limit of 150 gallons per day

(= 0.104 gpm) through any one SG. This LCO equates to a limit of 600 gallons per day

(= 0.417 gpm) total P-T-S leakage through all four SGs. The total P-T-S leakage is assumed to leak at a rate of 1 gpm into the faulted SG for 0-0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and then it continues to leak into the remaining three (3) intact SGs for 0.5-32 hours.

Primary-to-Secondary Leak - Primary Coolant Density Consistent with RG 1.183, Appendix F, Section 5.2, the primary coolant density used in converting the volumetric P-T-S leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) is assumed to be 1.0 gm/cc (62.4 lbm/ft3). Conversely, the temperature dependent specific volumes are used for converting the mass flow rates into volumetric flow rates to maximize the releases and consequently the doses.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Primary-to-Secondary Leak Duration Consistent with RG 1.183, Appendix F, Section 5.3, the P-T-S leakage is 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 212°F. In this analysis PC is conservatively assumed to leak into the intact SGs for 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months until the RHR system is initialized.

Release of Fission Products and Noble Gases Consistent with RG 1.183, Appendix F, Section 5.4, the release of fission products from the secondary system is evaluated with the assumption of coincident loss of offsite power (LOOP).

The offsite power is assumed to be lost so that the main steam condensers are not available for removal of the decay heat.

All noble gases radionuclides released from the primary system are assumed to be released to environment without reduction or mitigation.

Iodine and Particulate Transport Model Consistent with RG 1.183, Appendix F, Section 5.6, the transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates. The post-SGTR thermal hydraulic condition in the faulted SG is such that a large amount of PC released through the ruptured tube excessively increases the SG coolant mass inventory, which eliminates the possibility of SG dryout condition. The coolant mass in the intact SGs provides adequate tube submergence to eliminate the flashing of PC in the P-T-S leakage.

Iodine activity in the PC released from the ruptured tube that does not immediately flash and in the P-T-S leakage is released from the faulted SG to the environment in proportion to the steaming rate and the iodine partition coefficient value of 100 recommended in RG 1.183. This assumes that the PC that does not immediately flash mixes with the bulk water in the SG.

Iodine activity in the PC released from the ruptured tube that immediately flashes to vapor is assumed to be released without decontamination (i.e., no credit is taken for scrubbing as the vapor rises through the bulk water of the SG). The holdup time for the flashed vapor is negligible as the SG steam space is assumed to be 1 ft3 in the RADTRAD model, as a modeling simplification.

The flash fraction (FF) of the PC (from the rupture flow) is calculated using equation FF = (hf1 - hf2) / hfg.

Where hf1 is the enthalpy of PC at system operating temperature and pressure (i.e., for RCS),

hf2 is the enthalpy of liquid at saturation conditions (i.e., for the bulk (secondary) water in the SG), and hfg is the latent heat of vaporization for the liquid at the same saturation conditions.

The FF is conservative as a time-dependent calculation is not performed, and the initial enthalpy, hf1, for the PC (rupture flow) entering the faulted SG is based on RCS operating temperature and pressure (at the outlet, for High Tavg). Using the resulting flash fraction for the duration of the release into the faulted SG (i.e., 0 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ) overestimates the actual amount of PC that flashes to steam, as it will decrease with time (as RCS temperature, pressure, and break flow decrease).

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 For the P-T-S leakage iodine release from the three intact SGs, it is assumed that the SG tubes are submerged in the SGs liquid, and consequently the P-T-S leakage is assumed to be diluted within the SGs liquid. This is a valid assumption based on a study documented in WCAP-13132 as previously discussed. In addition, for the secondary liquid iodine release from the SGs, an iodine partition coefficient of 100 is used consistent with RG 1.183.

3.3.4 SGTR Atmospheric Dispersion Factors SGTR EAB /Q The EAB atmospheric dispersion factors are 1.97E-4 s/m3 and 2.44E-4 s/m3 for Units 1 and 2 SGTR accidents, respectively.

3.3.5 SGTR Key Analysis Assumptions and Inputs Method of Analysis RADTRAD (Reference 13) Version 3.03 is used to calculate the radiological consequences from airborne releases resulting from a SGTR. The analysis addresses the reactor coolant activity concentrations corresponding to (1) a pre-accident iodine spike and (2) a concurrent iodine spike. There is no fuel failure during this event. The SG liquid masses and steam mass releases are different for Salem Units 1 and 2 SGs. Therefore, the SGTR accident is analyzed separately for each unit.

Basic Data and Assumptions for SGTR The basic data and assumptions are summarized in Table 3.3-1.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.3-1 Basic Data and Assumptions for SGTR Parameter Value Used Source Term Data Licensed reactor core power level 3,459 MWt PC (iodine) specific activity in the Technical 1.0 µCi/gm DE I-131 Specifications PC (non-iodine) specific activity in the Technical 600 µCi/gm DE Xe-133 Specifications Maximum PC (iodine) specific activity permitted by the 60 µCi/gm DE Iodine-131 (for up to Technical Specifications 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after entering the LCO)

Concurrent iodine spiking factor 335 (for an SGTR accident)

Duration of concurrent iodine spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (for an SGTR accident)

Secondary coolant (iodine) specific activity in the 0.1 µCi/gm DE Iodine-131 Technical Specifications Maximum RCS volume 11,816 ft3 (Unit 1) 12,707 ft3 (Unit 2)

Total P-T-S leakage through faulted SG 1 gpm through faulted SG until (0 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ) isolated Total P-T-S leakage through all intact SGs 1 gpm (0.5 - 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months )

PC iodine activity concentration based on 1% failed fuel defects Activity (µCi/g) Activity Activity (µCi/g)

Isotope Isotope Isotope

(µCi/g)

I-131 2.8 I-133 4.2 I-135 2.3 I-132 2.8 I-134 0.57 PC noble gas activity based on 1% failed fuel defects Activity (µCi/g) Activity Activity (µCi/g)

Isotope Isotope Isotope

(µCi/g)

Kr-83m 4.0E-01 Kr-88 3.0E+00 Xe-135m 4.9E-01 Kr-85m 1.7E+00 Xe-131m 2.1E+00 Xe-135 8.5E+00 Kr-85 8.2E+00 Xe-133m 1.7E+01 Xe-138 6.1E-01 Kr-87 1.0E+00 Xe-133 2.6E+02 Maximum SG Dilution Mass 119,233 lbm (Unit 1) 127,312 lbm (Unit 2)

Fuel Damage No Fuel Failed Steam temperature 528.5°F (Unit 1) 531.4°F (Unit 2)

RCS average temperature 582.9°F (Units 1 & 2)

Iodine Half-lives Isotope Half-life (sec) Isotope Half-life Isotope Half-life (sec)

(sec)

I-131 0.694656E+06 I-133 0.7488E+05 I-135 0.23796E+05 I-132 0.828E+04 I-134 0.3156E+04 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Parameter Value Used Activity Transport Model PC released from RCS to faulted SG (i.e., 137,246.2 lb (Unit 1) rupture flow) 123,000 lb (Unit 2)

Steam mass released from faulted SG to the environment 56,500 lb (Unit 1) 0.0 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 65,000 lb (Unit 2)

PC leakage to intact SGs 1 gpm (0.5-32 hours) 0.191 cfm Primary-to-Secondary leak duration 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months Steam mass released from intact SGs to the environment Unit 1:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 465,000 lbs 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1,055,000 lbs 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1,503,000 lbs 24 - 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months 477,000 lbs 30 - 32.0 hours0 days 0 hours 0 weeks 0 months 150,333 lbs =

[(451,000) x (32 - 30)/(36 - 30)]

Unit 2:

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 530,000 lbs 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1,180,000 lbs 8 - 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months 2,050,000 lbs SG liquid Iodine partition coefficient 100 RCS nominal operating pressure 2235 psig (Units 1 & 2)

[2235 + 14.7 = 2249.7 psia]

RCS outlet temperature (High Tavg) 613.1°F (Units 1 & 2)

Design steam pressure (High Tavg) 869 psia (Unit 1) 900 psia (Unit 2)

Site Boundary Release Model Parameters EAB atmospheric dispersion factors (/Q) 1.97E-04 (Unit 1)

(sec/m3) 2.44E-04 (Unit 2)

EAB breathing rate (m3/s) 3.5E-04 EAB allowable dose limit Pre-Accident Iodine Spike Case 25 rem TEDE Concurrent Iodine Spike Case 2.5 rem TEDE 3.3.6 SGTR Analysis Results The total TEDE doses to the Control Room, EAB and LPZ from a SGTR accident are summarized in Table 3.3-2 through Table 3.3-5 for the concurrent and pre-accident spike for Units 1 and 2. The Unit 2 pre-accident spike case results in the highest dose consequences for the offsite assessment. All doses are within the limits specified in RG 1.183 and 10 CFR 50.67.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.3-2 Unit 1 SGTR Accident - Preaccident Iodine Spike Case TEDE (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 9.07E-03 3.20E-02 3.70E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

P-T-S Iodine Release (Flashing) 1.22E+00 2.10E+00 1.98E-01 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

SC Liquid Iodine Release 5.65E-04 7.49E-04 1.52E-04 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Noble Gas Release 9.34E-03 5.13E-02 4.91E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Total 1.24E+00 2.18E+00 2.07E-01 Allowable TEDE Limit 5.00E+00 2.50E+01 2.50E+01 Table 3.3-3 Unit 1 SGTR Accident - Concurrent Iodine Spike Case TEDE (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 2.46E-02 3.25E-02 7.02E-03 (occurs at t = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

P-T-S Iodine Release (Flashing) 3.68E-01 2.08E+00 1.96E-01 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

SC Liquid Iodine Release 5.65E-04 7.49E-04 1.52E-04 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Noble Gas Release 9.34E-03 5.13E-02 4.91E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Total 4.03E-01 2.16E+00 2.08E-01 Allowable TEDE Limit 5.00E+00 2.50E+00 2.50E+00 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.3-4 Unit 2 SGTR Accident - Preaccident Iodine Spike Case TEDE (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 9.00E-03 3.92E-02 3.66E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

P-T-S Iodine Release (Flashing) 9.94E-01 2.25E+00 1.72E-01 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

SC Liquid Iodine Release 6.20E-04 1.06E-03 1.67E-04 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Noble Gas Release 8.32E-03 5.83E-02 4.51E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Total 1.01E+00 2.35E+00 1.80E-01 Allowable TEDE Limit 5.00E+00 2.50E+01 2.50E+01 Table 3.3-5 Unit 2 SGTR Accident - Concurrent Iodine Spike Case TEDE (rem)

Receptor Location Control Room EAB LPZ P-T-S Iodine Release 2.26E-02 3.89E-02 6.34E-03 (occurs at t = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

P-T-S Iodine Release (Flashing) 2.96E-01 2.10E+00 1.60E-01 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

SC Liquid Iodine Release 6.20E-04 1.06E-03 1.67E-04 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Noble Gas Release 8.32E-03 5.83E-02 4.51E-03 (occurs at t = 0 hours0 days 0 hours 0 weeks 0 months )

Total 3.28E-01 2.20E+00 1.71E-01 Allowable TEDE Limit 5.00E+00 2.50E+00 2.50E+00 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.4 Other Affected Salem Analyses The revised EAB /Q value of 2.44E-4 s/m3 was also used to revise the dose consequence analyses for the other postulated events affected by the change to the EAB distance.

The current and revised calculated EAB dose values for the affected events are shown in Table 3.4-1 below, with the regulatory limit and the SRP-15.0.1 Table 1 accident criteria.

The increases in calculated EAB doses (1) are less than 10 percent of the difference between the current calculated dose values and the 10 CFR 50.67 regulatory limit, and (2) the increased doses do not exceed the current SRP guideline value for the applicable event. Therefore, the increases in calculated EAB doses are not considered more than minimal increases in the consequences of an accident evaluated in the UFSAR.

Table 3.4-1 Other Affected Salem Analyses Current Revised Change Regulatory SRP Current Revised Change EAB EAB in Event Limit Guideline Margin Margin in Unit Dose Dose Margin (rem (rem (rem (rem Margin (rem (rem (rem TEDE) TEDE) TEDE) TEDE) %

TEDE) TEDE) TEDE)

FHA in 1& 23.74 22.64 -1.1 -4.63 25 6.3 1.26 2.36 Containment 2 FHA in Fuel 1& 23.72 22.61 -1.11 -4.68 Handling 25 6.3 1.28 2.39 2

Building MSLB / Pre- 1 0.0674 0.127 24.9326 24.873 -0.0596 -0.24 Accident 25 25 Iodine Spike 2 0.0675 0.127 24.9325 24.873 -0.0595 -0.24 Case MSLB / 1 0.386 0.724 24.614 24.276 -0.338 -1.37 Concurrent 25 2.5 Iodine Spike 2 0.363 0.682 24.637 24.318 -0.319 -1.29 Case 1 0.115 0.216 24.885 24.784 -0.101 -0.41 LRA 25 2.5 2 0.107 0.201 24.893 24.799 -0.094 -0.38 1 0.251 0.472 24.749 24.528 -0.221 -0.89 REA 25 6.3 2 0.250 0.468 24.750 24.532 -0.218 -0.88 Although not discussed in RG 1.183, UFSAR Section 15.3.6 describes the evaluation of Volume Control Tank and Waste Gas Decay Tank ruptures. When updated to account for the revised EAB /Q, the resulting EAB 2-hour doses remain less than 0.1 rem TEDE.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.5 Hope Creek Loss of Coolant Accident (LOCA)

This section describes the methods employed and results obtained from the radiological reanalysis of the design basis LOCA for Hope Creek. The analysis considers dose contributions from the following post-LOCA release paths:

Containment Leakage

Engineered Safety Feature (ESF) Leakage

Main Steam Isolation Valve (MSIV) Bypass Leakage.

Doses are calculated at the EAB for the worst-case two-hour period. The methodology used to evaluate the EAB doses resulting from a LOCA is consistent with RG 1.183.

3.5.1 LOCA Scenario Description The design basis LOCA scenario for radiological calculations is initiated assuming a major rupture of the RCS piping. In order to yield radioactive releases of the magnitude specified in RG 1.183, 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, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 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 from there to the environment via paths listed above.

3.5.2 LOCA Source Term Definition RG 1.183 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 based on a maximum power level of 3917 MWt.

Table 3.5-1 lists the RG 1.183 source term inputs used in the LOCA analysis and includes: the core inventory release fractions by radionuclide group, timing of release, and chemical form of the release into Containment.

RG 1.183 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 in the subsequent 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Table 3.5-2 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 3.5-3 lists the isotopes and the associated power-normalized curies at the end of a fuel cycle that were input into RADTRAD. The CEDE and deep dose equivalent (DDE) dose conversion factors used for each of the isotopes were based on FGRs 11 and 12 (References 9 and 10 respectively).

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.5-1 RG 1.183 Source Term Input Data Characteristic RG 1.183 Source Term Noble Gases 100%

Iodine 30%

Cesium 25%

Core Fractions Released to Containment Tellurium 5%

Barium 2%

Others -0.02% to 0.25%

Timing of Release Released in Two Phases over 2-hour Interval 4.85% Elemental Iodine Chemical and Physical Form 0.15% Organic Vapor 95% Aerosol Solids Treated as an Aerosol Table 3.5-2 RG 1.183 Release Phases Core Release Fractions Early In-Isotope Group Gap Vessel Noble Gases 0.05 0.95 Halogens 0.05 0.25 Alkali Metals 0.05 0.20 Tellurium 0 0.05 Barium, Strontium 0 0.02 Noble Metals 0 0.0025 Cerium 0 0.0005 Lanthanides 0 0.0002 Duration (hours) 0.5 1.5 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.5-3 Core Inventory1 Isotope2 Ci/MWt Isotope2 Ci/MWt Isotope2 Ci/MWt Co-58 1.529E+02 Ru-103 8.000E+04 Cs-136 2.210E+03 Co-60 1.830E+02 Ru-105 2.900E+04 Cs-137 7.890E+03 Kr-85 3.760E+02 Ru-106 3.370E+04 Ba-139 4.920E+04 Kr-85m 6.980E+03 Rh-105 2.720E+04 Ba-140 4.730E+04 Kr-87 1.340E+04 Sb-127 2.940E+03 La-140 4.920E+04 Kr-88 1.890E+04 Sb-129 8.760E+03 La-141 4.410E+04 Rb-86 7.540E+01 Te-127 2.920E+03 La-142 4.330E+04 Sr-89 2.530E+04 Te-127m 3.890E+02 Ce-141 4.490E+04 Sr-90 3.000E+03 Te-129 8.710E+03 Ce-143 4.130E+04 Sr-91 5.060E+04 Te-129m 1.280E+03 Ce-144 7.485E+04 Sr-92 3.460E+04 Te-131m 2.801E+04 Pr-143 4.020E+04 Y-90 3.190E+03 Te-132 3.830E+04 Nd-147 1.800E+04 Y-91 3.280E+04 I-131 2.700E+04 Np-239 5.470E+05 Y-92 3.480E+04 I-132 3.900E+04 Pu-238 1.350E+02 Y-93 4.010E+04 I-133 5.480E+04 Pu-239 1.170E+01 Zr-95 4.760E+04 I-134 6.060E+04 Pu-240 1.690E+01 Zr-97 1.459E+05 I-135 6.260E+04 Pu-241 4.700E+03 Nb-95 4.780E+04 Xe-133 5.390E+04 Am-241 5.800E+00 Mo-99 5.130E+04 Xe-135 1.790E+04 Cm-242 1.510E+03 Tc-99m 4.470E+04 Cs-134 6.970E+03 Cm-244 1.020E+02 Notes:

1. Power level = 3,917 MWt

2. Noble Gases (Xe, Kr), Halogens (I), Alkali Metals (Cs, Rb), Tellurium Group (Te, Sb, Ba, Sr),

Noble Metals (Ru, Rh, Mo, Tc, Co), Lanthanides (La, Zr, Nd, Nb, Pr, Y, Cm, Am), Cerium (Ce, Pu, Np)

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.5.3 LOCA Atmospheric Dispersion Factors LOCA Offsite/EAB /Q The EAB atmospheric dispersion factor is 8.14E-4 sec/m3.

3.5.4 Containment Activity and Leakage Activity Transport in Primary Containment The average core inventory is released into the containment at the release timing and fractions shown in Table 3.5-1 and Table 3.5-2. Since the post-LOCA minimum suppression chamber water pH is greater than 7.0, the chemical form of radioiodine released into the containment is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Except for elemental and organic iodine and noble gases, the remaining fission products are assumed to be in particulate form. In accordance with Assumption 3.1 of Appendix A to RG 1.183, the radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (drywell). The radioactivity release into the containment is assumed to terminate at the end of the early in vessel phase, which occurs 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the onset of a LOCA.

Radioactivity is initially released from the core and diluted into the drywell air volume only.

Following the initial blowdown of the reactor pressure vessel, the steaming in the reactor pressure vessel (RPV) carries fission products to the containment. When core cooling is restored, the fuel damage is terminated. The steam and the ESF flows carry any remaining fission products from the vessel, through the break, to the primary containment and provide new steam flow for rapid drywell-suppression air space mixing. After 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the containment drywell and suppression chamber air volumes are expected to become well mixed due to a very high flow established between drywell and wetwell as a result of steaming and condensing phenomenon and thus after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , airborne activity dilution is credited in the combined air volumes of the drywell and suppression chamber. Taking credit for dilution of containment airborne activity within the drywell and suppression chamber combined air volumes over the remaining course of the accident (after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) is in accordance with guidance provided in Section 3.1 of Appendix A to RG 1.183. Confining the airborne activity to the drywell air volume during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is conservative because it increases the amount of airborne activity released from containment during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Reduction In Airborne Activity Inside Containment Aerosol Deposition The gravitational deposition of aerosols from the containment atmosphere is credited by using the time dependent aerosol natural deposition contamination coefficients with 10 percentile uncertainty distribution (resulting in the lowest removal rate of the aerosols from the containment). These deposition rates are used as user-defined coefficients in RADTRAD for the natural deposition mechanism in the containment compartment for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (i.e., no credit is taken for aerosol deposition after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). These deposition rates are applied to the post-LOCA containment and MSIV leakage release paths, which credit the aerosol deposition in the drywell.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Suppression Pool Scrubbing Iodine removal by suppression pool scrubbing is not credited because the timed release of activity associated with the AST methodology results in the bulk of core activity being released to containment well after the initial mass and energy release.

Natural Deposition of Elemental Iodine on Containment Surfaces In accordance with Section 3.2 of Appendix A to RG 1.183, reduction in containment airborne radioactivity by natural deposition within containment may be credited using the models described in SRP 6.5.2, even if no credit is taken for containment sprays. The Decontamination Factor (DF) of elemental iodine is based on the SRP 6.5.2 guidance and is limited to a DF of 200. The SRP 6.5.2 calculation of the elemental iodine removal rate is based on a minimized wetted surface area that conservatively results in a smaller elemental removal coefficient and a longer time to reach an elemental iodine DF of 200. This longer time to reach the DF allows the elemental iodine to remain airborne in the drywell atmosphere for a longer time prior to release to the environment via containment and MSIV leakage. This results in a larger release and a higher dose. The drywell wetted surface area is conservatively minimized by crediting 25% of the drywell lining surface, and 50% of the major equipment and structure surfaces, and then by applying a 25% reduction to the estimated surface area. The resultant modeled surface area of 33,200 ft2 is less than half of the available 69,126 ft2 drywell wetted surface area. The removal of elemental iodine by the wetted surface area is consistent with the guideline provided in SRP 6.5.2 and RG 1.183. The suppression pool water pH is greater than 7.0. A pH greater than 7 inhibits the iodine that deposits on containment surfaces from re-evolving back into containment atmosphere during containment spray recirculation.

RADTRAD is used to calculate the cutoff time of the elemental iodine removal by wall deposition inside the drywell. The elemental iodine DF of 200 is reached after 4.00 hrs. Therefore, elemental iodine wall removal is conservatively not credited in the analysis beyond 4.0 hours0 days 0 hours 0 weeks 0 months .

3.5.5 Containment Leakage The time dependent radionuclide source term airborne in containment and available for leakage credits dilution and reduction of airborne activity. The containment is assumed to leak at the TS leak rate of 0.5 volume percent per day over the course of the accident, i.e., 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months . No credit is taken for a reduction in containment leak rate (e.g., at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post LOCA or subsequently during the LOCA duration) due to a reduction in containment pressure with time.

Keeping the containment leakage rate at the TS limit for the duration of the accident is conservative since it maximizes the amount of radioactivity released which maximizes doses.

Primary containment leakage is into the reactor building; however, for the first 375 seconds primary containment leakage is released directly to the environment with no credit for holdup or filtration. The time of 375 seconds is the draw down time, i.e., the time it takes to bring the reactor building pressure down to 0.25-inch water gauge negative pressure relative to adjacent areas as defined in TS. After 375 seconds, primary containment leakage is via the reactor building through the Filtration, Recirculation and Ventilation System (FRVS) filter system. The FRVS recirculation system circulates reactor building air at a recirculation flow rate of 108,000 cfm. The volume of the reactor building is 4.00E+06 cubic feet and thus the FRVS recirculation system recirculates reactor building air at a rate of 1.62 reactor building volumes/hr. Although the FRVS provides good mixing of airborne activity within the reactor building, for conservatism, LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 credit is taken for mixing and dilution of activity in only 50% of the reactor building volume. To account for 50% mixing in the reactor building, the FRVS vent system exhaust rate to the environment is doubled. Airborne activity in the reactor building is reduced by both the FRVS recirculation and FRVS vent filtration systems before it is released to the environment.

Containment Purge during Normal Operation:

The use of the drywell and suppression chamber purge exhaust lines for pressure control during plant operational conditions 1, 2, and 3 is unrestricted provided 1) only the inboard purge exhaust isolation valves on these lines and the vent valves on the 2-inch vent paths are used and 2) the outboard purge exhaust isolation valves remain closed. The design of the purge supply and exhaust isolation valves and the 6 nitrogen supply valves meets the requirements of Branch Technical Position CSB 6-4, Containment Purging during Normal Plant Operations, that the maximum isolation time for each of these valves not exceed 5 seconds. Further, a dose analysis was performed that concludes that the dose associated with a containment purge while operating, concurrent with a LOCA, is negligible compared to the dose due to the release of design basis LOCA sources. The negligible purge dose is based on limiting the purge isolation valve closure time to 5 seconds and so the radiation source term released via the purge system is reactor coolant. Calculation of radiological consequences is therefore not required due to a normal containment purge coincident with a LOCA.

Containment Purge during LOCA:

Section 7 of Appendix A to RG 1.183 states that if post LOCA primary containment purging is performed as a combustible gas or pressure control measure or if primary containment purging is required within 30 days following a LOCA, then radiological consequences should be analyzed. It further states that if the containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, then radiological consequences need not be evaluated. The Hope Creek containment is not purged for combustible gas or pressure control measure within 30 days following a LOCA.

Therefore, in accordance with guidance provided in Section 7 of Appendix A to RG 1.183, radiological consequences associated with a purge release during LOCA conditions are not evaluated.

3.5.6 Engineered Safety Feature (ESF) System Leakage The ESF systems that recirculate suppression pool water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. The radiological consequences from the postulated leakage are analyzed and combined with the consequences from other fission product release paths to determine the total calculated radiological consequences from the LOCA. The ESF components are located in the Reactor Building.

The ESF leakage rate is 2.85 gpm. This value is doubled to 5.70 gpm and assumed to start at time t=0.0 minutes after onset of a LOCA. Ten percent of the iodine in the suppression pool water that leaks becomes airborne. All remaining fission products in the recirculating liquid are assumed to be retained in the liquid phase.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 ESF leakage is into the reactor building. As is the case for containment leakage, airborne radioactivity released due to ESF leakage is released directly to the environment for the first 375 seconds (i.e., during draw down). After 375 seconds, credit is taken for mixing in the reactor building volume and for airborne activity removal by the FRVS recirculation and exhaust filter systems.

Suppression Pool Water Source Term:

Except for noble gases, all the fission products released from the fuel to the containment are assumed to instantaneously and homogeneously mix in the suppression pool water at the time of release from the core. The radioiodine that is postulated to be available for release to the environment is assumed to be 97% elemental and 3% organic.

3.5.7 Post-LOCA MSIV Leakage Pathway The four main steam lines, which penetrate the primary containment, are automatically isolated by the MSIVs in the event of a LOCA. There are two MSIVs on each steam line, one inside containment and one outside containment. The MSIVs are functionally part of the primary containment boundary and design leakage through these valves provides a leakage path for fission products that bypass the secondary containment and enter the environment as a ground-level release. It is assumed that one steam line ruptures between the RPV and the inboard MSIV. The line with the rupture is referred to as the failed steam line. In addition, it is assumed that the inboard MSIV fails open on the failed steam line. MSIV leakage is postulated to be released to the environment through the failed steam line and one of the three remaining intact steam lines.

The MSIVs are postulated to leak at a total design leak rate of 250 scfh at 50.6 psig. This is the maximum allowable TS value (250 scfh combined through all 4 main steam lines). Measured leakage through any one steam line cannot exceed 150 scfh. In order to maximize dose results, the leak rate of 150 scfh is applied to the failed steam line with its failed MSIV and the remainder leak rate of 100 scfh (250 scfh total - 150 scfh maximum for one line) is applied to a single intact main steam line. Use of the maximum TS leak rate is in accordance with Assumption 6.2 of Appendix A to RG 1.183.

The time dependent radionuclide source term airborne in containment and available for leakage via the main steam lines is adjusted for dilution and reduction of airborne activity. No credit is taken for a reduction in containment leak rate due to a reduction in containment pressure. This is also true for the MSIV leak rate (i.e., no credit is taken for a reduction in the MSIV leak rate due to a reduction in containment pressure at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post LOCA or subsequently during the LOCA duration).

Note that credit is taken for deposition of aerosol activity in the main steam lines due to gravitational deposition with 40th percentile aerosol settling velocity in piping segments downstream of the outboard MSIV in the failed and intact main steam lines. A reduction in elemental iodine activity is credited based on natural removal efficiencies. Deposition of aerosol and elemental iodine is assumed for 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months post-LOCA, even though the release continues out to 30 days. The modeling used for deposition and plateout in the main steam lines was previously reviewed by the NRC in Hope Creek License Amendment No. 184 (Reference 19) and determined to be acceptable.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.5.8 Data for LOCA Model Table 3.5-4 provides a summary of parameters used to reanalyze the EAB LOCA dose analysis for Hope Creek.

Table 3.5-4 Data for LOCA Model Parameter Value Used Activity Transport in Primary Containment Primary Containment Parameters Drywell Air Volume 169,000 ft3 Suppression Chamber Air Volume 137,000 ft3 Containment Air Volume 306,000 ft3 Containment Leak Rate 0-720 hrs 0.5 v%/day Draw Down Time 375 sec Cont. Leakage Before Draw Down Time Directly Released to Environment

(< 375 sec)

Cont. Leakage After Draw Down Time Directly Released to Reactor Building

(>375 sec)

Reactor Building Volume 4,000,000 ft3 Reactor Building Mixing 50%

FRVS Vent Exhaust Rate Before Draw 9000 cfm +/- 10%

Down FRVS Vent Exhaust Flow Rate After 3324 + 5676e-1.18t Draw Down [Note 1]

FRVS Vent Exhaust Filter Efficiency Iodine Species Efficiency (%)

Elemental 90%

Aerosol 99%

Organic 90%

FRVS Recirculation Filter Efficiency Iodine Species Efficiency (%)

Elemental 0%

Aerosol 99%

Organic 0%

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Parameter Value Used Post Draw Down FRVS Exhaust Rates For 50% Mixing Post-LOCA Time (hr) Normal Flow Rate 50% Mixing Flow Rate (cfm) (cfm)

A = 3324 + 5676e-1.18t A x 1.1 x 2

[Note 2]

0 9000 19800 0.104 (375 sec) 9000 19800 0.437 7154 15739 2.104 3860 8492 4.104 3375 7425 8.104 3324 7313 24 3324 7313 96 3324 7313 FRVS Recirculation Flow Rate 120,000 cfm - 10% [Note 2]

(or, 108,000 cfm)

ESF Leakage Model Parameters Suppression Pool Water Volume 118,000 ft3 ESF Leakage 2.85 gpm ESF Leakage Initiation Time 0 minute Suppression Pool Water pH >7 Suppression Pool Water Activity Group Gap Release Early In-Vessel Release Phase Phase Timing Duration (Hrs) 2 min - 0.50 0.50 - 2.0 hours0 days 0 hours 0 weeks 0 months hours Halogen 0.05 0.25 Iodine Flashing Factor 10%

Chemical Form Iodine In ESF Leakage Elemental 97%

Organic 3%

Pool Peak Temperature 212.30F LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Parameter Value Used MSIV Leakage Model Parameters Total MSIV Leak Rate Through All Four 250 scfh Lines MSIV Leak Rate Through Line With 150 scfh MSIV Failed MSIV Leak Rate Through First Intact 100 scfh Line Limiting Steam Line Temperature 546°F 550°F Number of Steam Lines 4 Diameter and Wall Thickness of Pipe Diameter = 26 Between RPV Nozzle and Inboard Wall Thickness = 1.117 Isolation Valves HV F022A/B/C/D Diameter and Wall Thickness of Pipe Diameter = 26 Between Inboard and Outboard Isolation Wall Thickness = 1.117 Valves HV F028A/B/C/D Diameter and Wall Thickness of Pipe Diameter = 26 Between Outboard and 3rd Isolation Wall Thickness = 1.023 Valves HV 3631A/B/C/D Diameter of Pipe Between 3rd Isolation Diameter = 28 and Turbine Stop Valves MSV1/2/3/4 Wall Thickness = 0.934 Corrosion Allowance For Steam 0.12 Drywell Peak Pressure 50.6 psig Drywell Peak Temperature 2980F Site Boundary Release Model Parameters EAB /Q (0-2 Hrs) 8.14E-04 sec/m3 EAB Breathing Rate 3.5E-04 m3/sec Note 1: Analytical expression for FRVS vent fan exhaust rate is from a plant-specific calculation incorporating as-built heat loading.

Note 2: Use of higher FRVS vent flow and lower FRVS recirculation flow is conservative.

3.5.9 Hope Creek LOCA Results Table 3.5-5 lists the TEDE dose at the EAB from a LOCA at Hope Creek. The dose at the EAB is less than the 25 rem TEDE limit stated in 10 CFR 50.67 and RG 1.183. The EAB dose represents the worst 2-hour dose for each release pathway.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.5-5 EAB Dose Summary for a Hope Creek LOCA Post-LOCA Activity EAB Release Path Post-LOCA TEDE (Rem)

Containment Leakage 1.58E+00 (occurs at t = 3.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

ESF Leakage 2.25E+00 (occurs at t = 10.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months )

MSIV Leakage 9.11E+00 (occurs at t = 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months )

Total 1.29E+01 Allowable TEDE Limit 2.50E+01 LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.6 Hope Creek Main Steam Line Break This section describes the methods employed and EAB dose results for the MSLB design basis radiological analysis. This analysis includes doses associated with the releases of radioactive material initially present in reactor coolant and reactor steam at maximum allowable TS concentrations for a pre-existing iodine spiking scenario and a maximum reactor coolant equilibrium iodine concentration scenario. No fuel failure is expected. The methodology used to evaluate the dose consequences resulting from the MSLB is consistent with RG 1.183.

3.6.1 MSLB Scenario Description The MSLB accident begins with a break in one of the main steam lines outside containment.

The break results in activity release to the environment. The total mass of coolant released is assumed to be the amount in the steam line and connecting lines at time of break plus the amount that passes through the valves prior to closure. All radioactivity in the released coolant is assumed to be released to the atmosphere instantaneously as a ground-level release.

Potential release paths are the blow out panels, south plant vent, and turbine building louvers.

Since the MSLB accident is a high-energy line break accident, the pressure sensitive blow out panels would break open immediately to relieve the high-pressure steam release 3.6.2 MSLB Source Term Definition The analysis of the MSLB accident assumes that no fuel rod failures occur as a result of the transient. Since no or minimal fuel damage is postulated for the limiting event, the released activity is the maximum coolant activity allowed by the TS. Two cases are considered, i.e., a pre-accident iodine spike case and a maximum reactor coolant equilibrium iodine concentration case.

Pre-Accident Iodine Spike Source Term The reactor coolant activity concentration for this case is assumed to be at the maximum TS value of 4.0 µCi/gm Dose Equivalent (DE) I-131 permitted for a condition of a pre-accident iodine spike. The iodine scaling factors for the pre-accident iodine spike case are based on the maximum iodine concentration of 4.0 µCi/gm using the following definition of DE I-131:

DOSE EQUIVALENT I-131 shall be that concentration of I-131 µCi/gm, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The iodine isotopic dose conversion factors are calculated using the thyroid dose conversion factors from FGR 11. The isotopic noble gas concentrations are calculated using the noble gas release rate at time t = 0 sec and the associated steam mass flow rate.

The use of the RADTRAD code requires a volume node for the source activity released from a MSLB accident. Therefore, a source volume of 200,000 ft3 is introduced for a MSLB accident release in a way that all activities are released to the environment in a single puff with a release rate of 2.0E+06 volume percent /day (i.e., equal to 20,000 volumes/day, and equal to approximately 14 volumes/minute). A reactor coolant mass of 140,000 pounds is assumed to be release from the MSLB. Although this release consists of two-phase flow of water and steam mixture with different iodine concentrations in each phase, it is conservatively assumed that the LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 reactor coolant iodine concentrations are appropriate for both phases. Similarly, the noble gas concentrations are assumed equal for both phases.

Equilibrium Condition Source Term The reactor coolant concentration for this case is assumed to be at a TS value of 0.2 µCi/gm DE I-131 permitted for an equilibrium iodine activity for continued full power operation. The specific release model, assumptions and design input parameters used in the analysis are the same as those used for the pre-accident iodine case except the isotopic iodine concentrations are calculated based on 0.2 µCi/gm DE I-131 and noble gas isotopic concentrations. The iodine scaling factor for the equilibrium iodine concentration case is based on the maximum iodine concentrations of 0.2 µCi/gm using the definition of DE I-131 that was used for the pre-accident iodine spike case.

3.6.3 MSLB Atmospheric Dispersion Factors MSLB EAB /Q The EAB atmospheric dispersion factor is 8.14E-4 s/m3.

3.6.4 MSLB Key Analysis Assumptions and Inputs Method of Analysis RADTRAD (Reference 13) Version 3.03 is used to calculate the radiological consequences from airborne releases resulting from a MSLB. This analysis addresses the reactor coolant activity concentrations corresponding to (1) a pre-accident iodine spike and (2) a maximum reactor coolant equilibrium iodine concentration. There is no potential fuel failure during this event.

Basic Data and Assumptions for MSLB The basic data and assumptions are summarized in Table 3.6-1.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Table 3.6-1 Basic Data and Assumptions for MSLB Parameter Value Used Source Term Rated thermal power 4,031 MWt Design Basis Iodine Coolant Concentration (µCi/gm)

Isotope Activity Isotope Activity Isotope Activity I-131 1.30E-02 I-132 1.20E-01 I-133 8.90E-02 I-134 2.40E-01 I-135 1.30E-01 Emission Rate of Noble Gases @ time t = 0 (µCi/sec)

Kr-83m 3.40E+03 Kr-88 2.00E+04 Xe-135m 2.60E+04 Kr-85m 6.10E+03 Xe-131m 1.50E+01 Xe-135 2.20E+04 Kr-85 2.00E+01 Xe-133m 2.90E+02 Xe-138 8.90E+04 Kr-87 2.00E+04 Xe-133 8.20E+03 Maximum reactor coolant iodine 4.0 µCi/gm concentration for pre-accident spike Maximum equilibrium reactor coolant iodine 0.2 µCi/gm concentration for continued full power operation Mass of reactor coolant released from MSLB 140,000 lbs accident Activity Transport Activity release rate 2.0E+06 percent source volumes/day; Assumed to postulate a single puff Duration of release Instantaneously in a single puff Type of release to the atmosphere Ground level release Chemical form of Iodine in reactor coolant released from the main steam line Aerosol 95%

Elemental 4.85%

Organic 0.15%

Source volume 200,000 ft ; Assumed to facilitate RADTRAD 3

nodalization Steam mass flow rate 17,774,000 lb/hr Site Boundary Release Model Parameters EAB atmospheric dispersion factor (/Q) 8.14E-04 (sec/m3)

EAB breathing rate (m3/sec) 3.5E-04 Allowable dose limits for pre-accident iodine spike 25 rem TEDE EAB Allowable dose limits for equilibrium condition 2.5 rem TEDE EAB LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.6.5 MSLB Analysis Results The total TEDE at the EAB from a MSLB is summarized below in Table 3.6-2 for the maximum reactor coolant equilibrium iodine concentration condition and the pre-accident iodine spike.

The doses are within the limits specified in RG 1.183 and 10 CFR 50.67.

Table 3.6-2 Dose Summary for the MSLB Accident Location TEDE (rem) Limit (rem)

Maximum Equilibrium Iodine Concentration Condition 2.41E-01 EAB (occurs at t = 0 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months )

Pre-Accident Iodine Spike 4.05E+00 EAB (occurs at t = 0 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months )

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 3.7 Other Affected Hope Creek Analyses The revised EAB /Q value of 8.14E-4 sec/m3 was also used to revise the dose consequence analyses for the other postulated events affected by the change to the EAB distance.

The current and revised calculated EAB dose values for the affected events are shown in Table 3.7-1 below, with the regulatory limit and the SRP-15.0.1 Table 1 accident criteria.

The increases in calculated EAB doses (1) are less than 10 percent of the difference between the current calculated dose values and the 10 CFR 50.67 regulatory limit, and (2) the increased doses do not exceed the current SRP guideline value for the applicable event. Therefore, the increases in calculated EAB doses are not considered more than minimal increases in the consequences of an accident evaluated in the UFSAR.

Table 3.7-1 Other Affected Hope Creek Analyses Current Revised Change Regulatory SRP Current Revised Change EAB EAB in Limit Guideline Margin Margin in Event Dose Dose Margin (rem (rem (rem (rem Margin (rem (rem (rem TEDE) TEDE) TEDE) TEDE) (%)

TEDE) TEDE) TEDE)

CRDA 25 6.3 0.112 0.480 24.888 24.52 -0.368 -1.48 FHA 25 6.3 0.533 2.29 24.467 22.71 -1.757 -7.18 2.5 FWLB 25 0.00287 0.0123 24.99713 24.9877 -0.00943 -0.04

[Note 1]

2.5 ILPB 25 0.074 0.316 24.926 24.684 -0.242 -0.97

[Note 1]

Note 1: Assumed to be 10% of Regulatory Dose Limit to be consistent with guidance in Standard Review Plan (SRP) 15.0.1.

Although not discussed in RG 1.183, UFSAR Section 15.7.1 describes the evaluation of an off-gas-treatment system failure based on the guidance provided in Branch Technical Position (BTP) ETSB 11-5, Rev. 0 (Reference 14). When adjusted to account for the revised EAB /Q, the resulting dose remains within the BTP criteria of 500 mrem individual exposure for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at the nearest exclusion area boundary.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 4 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.2 defines the exclusion area as that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. PSEG continues to have the required authority and control over the Exclusion Area.

10 CFR 50.67, "Accident source term," establishes acceptable radiation dose limits resulting from design basis accidents for an individual located at the exclusion area boundary or low population zone, and for occupants of the control room. The revised design basis radiological analyses incorporating the proposed changes to the Exclusion Area Boundary demonstrate the regulatory limits in 10 CFR 50.67 continue to be met.

Regulatory Guide 1.23, "Meteorological Monitoring Programs for Nuclear Power Plants,"

provides guidance on the measurement and processing of onsite meteorological data for use as input to atmospheric dispersion models in support of plant licensing and operation.

Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," identifies acceptable methods for calculating offsite atmospheric relative concentration (/Q) values. The revised EAB /Q values were calculated using methodology consistent with the guidance in Regulatory Guide 1.145.

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

NUREG-0800, SRP Section 15.0.1, "Radiological Consequences of Analyses Using Alternative Source Terms," dated July 2000, provides guidance for an application for the initial implementation of an alternative source term at operating power reactors and subsequent LARs from these plants. The plant-specific information provided in this License Amendment Request addresses the guidance in SRP 15.0.1.

Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," provides guidance to ensure that the appropriate level of technical detail is considered in AST analyses and included in AST submittals.

4.2 No Significant Hazards Consideration Determination Analysis In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) requests amendments to Renewed Facility Operating License Nos. DPR-70 and DPR-75 for Salem Generating Station Units 1 and 2 (Salem) and NPF-57 for Hope Creek Generating Station (Hope Creek). The proposed amendments would change the licensing basis as described in the Salem and Hope Creek Updated Final Safety Analysis Reports (UFSARs) to account for modifications to the Exclusion Area Boundary (EAB) for Salem and Hope Creek. Revised dose consequence calculations performed in accordance with Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Plants: LWR Edition," (SRP) Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms," demonstrate the radiological consequences for postulated events remain within regulatory limits.

PSEG has evaluated whether a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

Updating the radiological consequence analyses to account for modifications to the Exclusion Area Boundary does not require any changes to any plant structures, systems, or components (SSCs) and therefore does not affect any accident initiators. As a result, the proposed changes do not significantly increase the probability of an accident. The revised assessments of the radiological consequences due to design basis accidents listed in the Salem and Hope Creek UFSARs conclude that the dose consequences affected by the proposed changes remain within the limits of 10 CFR 50.67 and within the limits of Regulatory Guide 1.183. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

Updating the radiological consequence analyses to account for modifications to the Exclusion Area Boundary does not change the design function or operation of any SSCs and does not introduce any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

No design basis or safety limits are exceeded or altered by this change. The revised assessments of the radiological consequences due to design basis accidents listed in the Salem and Hope Creek UFSARs conclude that the dose consequences affected by the proposed changes remain within the limits of 10 CFR 50.67 and within the limits of Regulatory Guide 1.183.

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

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 Based on the above, PSEG concludes that the proposed amendment does not involve a 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.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents 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 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.

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02 6 REFERENCES

1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

2. NRC Letter to PSEG, "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendments Re: Alternate Source Term (TAC Nos. MC3094 and MC3095)," dated February 17, 2006 (ADAMS Accession No. ML060040322)

3. NRC Letter to PSEG, "Hope Creek Generating Station - Issuance of Amendment Re:

Increase in Allowable Main Steam Isolation Valve (MSIV) Leakage Rate and Elimination of MSIV Sealing System (TAC No. MB1970)," dated October 3, 2001 (ADAMS Accession No. ML012600176)

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

5. Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Revision 1, November 1982

6. Regulatory Guide 1.23, "Meteorological Monitoring Programs for Nuclear Power Plants,"

Revision 1, March 2007

7. NUREG-2202 "Safety Evaluation Report - Related to the Early Site Permit Application in the Matter of PSEG Power, LLC and PSEG Nuclear, LLC for the PSEG Early Site Permit Site," October 2016 (ADAMS Accession No. ML16280A024)

8. Site Safety Analysis Report, Rev. 4, PSEG Site Early Site Permit Application, Date Released: June 24, 2015, Accession Number: ML15169A740.

9. Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentrations and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, EPA 520/1-88-020, Environmental 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, 1992

11. NUREG-0800, Standard Review Plan 6.5.2, "Containment Spray as a Fission Product Cleanup System," Rev. 2, 1988

12. Westinghouse Study WCAP-13132, "The Effect of Steam Generator Tube Uncovery on Radioiodine Release," January 1992

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

14. Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," Rev. 0, July 1981

15. NRC Letter to PSEG, "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendment Re: Refueling Operations - Fuel Decay Time Prior To Commencing Core Alterations or Movement of Irradiated Fuel (TAC Nos. MB5488 and MB5489)," dated October 10, 2002 (ADAMS Accession No. ML022770181)

16. NRC Letter to PSEG, "Hope Creek Generating Station - Issuance of Amendment Re:

Containment Requirements during Fuel Handling and Removal of Charcoal Filters (TAC No. MB5548)," dated April 15, 2003 (ADAMS Accession No. ML030760293)

17. NRC Letter to PSEG, "Hope Creek Generating Station - Issuance of Amendment Re:

Extended Power Uprate (TAC No. MD3002)," dated May 14, 2008 (ADAMS Accession No. ML081230540)

LR-N23-0050 LAR S23-04 Enclosure LAR H23-02

18. NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," March 7, 2006

19. NRC letter to PSEG, Hope Creek Generating Station - Issuance of Amendment 184 Re: Use of Isotope Test Assemblies For Cobalt-60 Production (TAC No. ME2949),

dated October 7, 2010 (ADAMS Accession No. ML102700263).

ML23249A261

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2.1 Background

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