Text

and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendments Nos. 230, 342, and 323 Delete Definition in 10 CFR 20 and Figures of Site and Surrounding Areas
	ML22012A435
Person / Time
Site:	Salem, Hope Creek
Issue date:	02/14/2022
From:	James Kim; Plant Licensing Branch 1
To:	Carr E; Public Service Enterprise Group
	Kim J
References
	EPID L-2021-LLA-0171
	Download: ML22012A435 (58)
	v • d • e

February 14, 2022 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION AND SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 230, 342, AND 323 RE: DELETION OF SPECIFIED DEFINITIONS AND FIGURES (EPID L-2021-LLA-0171)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 230, 342, and 323 to Renewed Facility Operating License Nos. NPF-57, DPR-70, and DPR-75 for the Hope Creek Generating Station (Hope Creek), and Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), respectively, in response to your application dated September 29, 2021, as supplemented by letter dated November 18, 2021.

The amendments revise the Hope Creek and Salem technical specifications (TSs) to remove TS definitions for Member(s) of the Public, Site Boundary, and Unrestricted Area which are already present in the definitions found in the Offsite Dose Calculation Manual for each site as well as Title 10 of the Code of Federal Regulations, Part 20, Section 1003. The amendments also remove figures of the site and surrounding area from the TSs.

E. Carr A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-354, 50-272, and 50-311

Enclosures:

1. Amendment No. 230 to NPF-57

2. Amendment No. 342 to DPR-70

3. Amendment No. 323 to DPR-75

4. Safety Evaluation cc: Listserv

PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 230 Renewed License No. NPF-57

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, dated September 29, 2021, as supplemented by letter dated November 18, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-57 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 230, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2022.02.14 Danna 11:28:28 -05'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 14, 2022

ATTACHMENT TO LICENSE AMENDMENT NO. 230 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of Renewed Facility Operating License No. NPF-57 with the attached revised pages. The revised pages are identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert i i ii ii iii iii xxiii xxiii 1-4 1-4 1-7 1-7 1-9 1-9 5-1 5-1 5-2 5-2 5-3 5-3 6-16b 6-16b 6-16c 6-16c 6-16d 6-16d

reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core Power levels not in excess of 3902 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 230, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-57 Amendment No. 230

DEFINITIONS SECTION 1.0 DEFINITIONS ................................................................................................................................ PAGE 1.1 ACTION ............................................................................................................................................... 1-1 1.2 DELETED ............................................................................................................................................ 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ............................................................... 1-1 1.4 CHANNEL CALIBRATION .................................................................................................................. 1-1 1.5 CHANNEL CHECK ............................................................................................................................. 1-1 1.6 CHANNEL FUNCTIONAL TEST ......................................................................................................... 1-1 1.7 CORE ALTERATION .......................................................................................................................... 1-2 1.8 DELETED ............................................................................................................................................ 1-2 1.9 CORE OPERATING LIMITS REPORT ............................................................................................... 1-2 1.10 CRITICAL POWER RATIO ................................................................................................................ 1-2 1.11 DOSE EQUIVALENT I-131 ................................................................................................................ 1-2 1.11.1 DRAIN TIME ................................................................................................................................. 1-2a 1.12 E-AVERAGE DISINTEGRATION ENERGY .................................................................................... 1-2a 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........................................... 1-2 1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME .............................. 1-3 1.15 DELETED ........................................................................................................................................... 1-3 1.16 DELETED ........................................................................................................................................... 1-3 1.17 FREQUENCY NOTATION ................................................................................................................. 1-3 1.18 IDENTIFIED LEAKAGE ..................................................................................................................... 1-3 1.18.1 INSERVICE TESTING PROGRAM ................................................................................................ 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME ......................................................................................... 1-3 1.20 LIMITING CONTROL ROD PATTERN .............................................................................................. 1-3 1.21 LINEAR HEAT GENERATION RATE ................................................................................................ 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST .............................................................................................. 1-4 1.23 DELETED ........................................................................................................................................... 1-4 1.24 Not Used 1.25 MINIMUM CRITICAL POWER RATIO ............................................................................................... 1-4 HOPE CREEK i Amendment No. 230

INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.26 OFF-GAS RADWASTE TREATMENT SYSTEM.................................................................. 1-4 1.27 OFFSITE DOSE CALCULATION MANUAL .........................................................................1-4 1.28 OPERABLE - OPERABILITY.................................................................................................1-5 1.29 OPERATIONAL CONDITION - CONDITION........................................................................ 1-5 1.30 PHYSICS TESTS ..................................................................................................................1-5 1.30-1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)..........................................1-5 1.31 PRESSURE BOUNDARY LEAKAGE ...................................................................................1-5 1.32 PRIMARY CONTAINMENT INTEGRITY ..............................................................................1-5 1.33 PROCESS CONTROL PROGRAM ......................................................................................1-6 1.34 PURGE-PURGING ...............................................................................................................1-6 1.35 RATED THERMAL POWER .................................................................................................1-6 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME ..................................................... 1-6 1.37 REPORTABLE EVENT .........................................................................................................1-6 1.38 ROD DENSITY .....................................................................................................................1-6 1.39 SECONDARY CONTAINMENT INTEGRITY ....................................................................... 1-7 1.40 SHUTDOWN MARGIN ......................................................................................................... 1-7 1.41 Not Used 1.42 Not Used ...............................................................................................................................1-8 1.43 SOURCE CHECK .................................................................................................................1-8 1.44 SPIRAL RELOAD .................................................................................................................1-8 1.45 SPIRAL UNLOAD .................................................................................................................1-8 1.46 STAGGERED TEST BASIS ................................................................................................. 1-8 1.47 THERMAL POWER ..............................................................................................................1-8 1.48 TURBINE BYPASS SYSTEM RESPONSE TIME ................................................................1-9 HOPE CREEK ii Amendment No. 230

INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.49 UNIDENTIFIED LEAKAGE ................................... ....... .................................. 1-9 1.50 Not Used 1.51 VENTILATION EXHAUST TREATMENT SYSTEM .......... .............................. 1-9 1.52 VENTING............. ......................... ................................................................. 1-9 TABLE 1.1, SURVEILLANCE FREQUENCY NOTATION........ .................................. 1-10 TABLE 1.2, OPERATIONAL CONDITIONS ........ ..................................... ................ 1-11 HOPE CREEK iii Amendment No. 230

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE LOCATION ......................... ....................................................................... 5-1 5.2 CONTAINMENT Configuration .............................................................................................. 5-1 Design Temperature and Pressure ............................................................. 5-1 Secondary Containment ............................................................................. 5-1 5.3 REACTOR CORE Fuel Assemblies......................... ................................................................ . 5-4 Control Rod Assemblies ............................................................................. 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ............................................................. 5-4 Volume ....................................................................................................... 5-5 5.5 METEOROLOGICAL TOWER LOCATION ........................................................... 5-5 5.6 FUEL STORAGE Criticality ..................................................................................................... 5-5 Drainage ..................................................................................................... 5-5 Capacity ..................................................................................................... 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT.................................................... 5-5 Table 5.7.1 1 Component Cyclic or Transient Limits ....................... 5-6 HOPE CREEK xxiii Amendment No. 230

DEFINITIONS LIMITING CONTROL ROD PATTERN 1.20 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

1.23 DELETED 1.24 Not Used MINIMUM CRITICAL POWER RATIO 1.25 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFF-GAS RADWASTE TREATMENT SYSTEM 1.26 An OFF-GAS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting reactor coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report required by Specifications 6.9.1.6 and 6.9.1.7.

HOPE CREEK 1-4 Amendment No. 230

DEFINITIONS SECONDARY CONTAINMENT INTEGRITY 1.39 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve or damper, as applicable secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b. All secondary containment hatches and blowout panels are closed and sealed.

c. The filtration, recirculation and ventilation system is in compliance with the requirements of Specification 3.6.5.3.

d. For double door arrangements, at least one door in each access to the secondary containment is closed, except when the access opening is being used for entry and exit.

e. For single door arrangements, the door in each access to the secondary containment is closed, except for normal entry and exit.

f. The sealing mechanism associated with each secondary containment penetration, e.g., welds, bellows or O-rings, is OPERABLE.

g. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a, except as indicated by the footnote for Specification 4.6.5.1.a.

SHUTDOWN MARGIN (SDM) 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

1.41 Not Used HOPE CREEK 1-7 Amendment No. 230

DEFINITIONS TURBINE BYPASS SYSTEM RESPONSE TIME 1.48 The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two separate time intervals: a) time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established, and b) the time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve. Either response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.

UNIDENTIFIED LEAKAGE 1.49 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

1.50 Not Used VENTILATION EXHAUST TREATMENT SYSTEM 1.51 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.52 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

HOPE CREEK 1-9 Amendment No. 230

5.0 DESIGN FEATURES 5.1 SITE LOCATION Hope Creek Generating Station is located in Salem County, New Jersey along the eastern shore of the Delaware River approximately 8 miles southwest of Salem, New Jersey and 18 miles south of Wilmington, Delaware.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a spherical lower portion, a cylindrical middle portion, and a hemispherical top head which form a drywell. The drywell is attached to the suppression chamber through a series of downcomer vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The drywell has a nominal free air volume of 169,000 cubic feet. The suppression chamber has an air volume of 137,000 cubic feet and a water region as described in Technical Specification Bases 3/4.6.2, Depressurization Systems.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum internal pressure 62 psig.

b. Maximum internal temperature: drywell 340oF.

suppression pool 310oF.

c. Maximum external differential pressure 3 psid.

SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Reactor Building, and a portion of the main steam tunnel and has a free volume of 4,000,000 cubic feet.

HOPE CREEK 5-1 Amendment No. 230

Intentionally Left Blank HOPE CREEK 5-2 Amendment No. 230

Intentionally Left Blank HOPE CREEK 5-3 Amendment No. 230

ADMINISTRATIVE CONTROLS 6.8.4.f Primary Containment Leakage Rate Testing Program A program shall be established, implemented, and maintained to comply with the leakage rate testing of the containment as required by 10CFR50.54(o) and 10CFR50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 50.6 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.5% of primary containment air weight per day.

Leakage Rate Acceptance Criteria are:

a. Primary containment leakage rate acceptance criterion is less than or equal to 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa,

2) Door seal leakage rate less than or equal to 5 scf per hour when the gap between the door seals is pressurized to greater than or equal to 10.0 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

6.8.4.g Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to member(s) of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

HOPE CREEK 6-16b Amendment No. 230

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2) Limitations on the concentration of radioactive material released in liquid effluents to unrestricted areas conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, HOPE CREEK 6-16c Amendment No. 230

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4.g Radioactive Effluent Controls Program

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

10) Limitations on venting and purging of the containment through the Reactor Building Ventilation System, Hardened Torus Vent, or the FRVS to maintain releases as low as reasonably achievable, and

11) Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

h. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluents monitoring program and modeling of the environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census, and

3) Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

HOPE CREEK 6-16d Amendment No. 230

PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 342 Renewed License No. DPR-70

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees), dated September 29, 2021, as supplemented by letter dated November 18, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-70 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 342, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2022.02.14 Danna 11:29:07 -05'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: February 14, 2022

ATTACHMENT TO LICENSE AMENDMENT NO. 342 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Replace the following page of Renewed Facility Operating License No. DPR-70 with the attached revised page as indicated. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of Renewed Facility Operating License No. DPR-70 with the attached revised pages as indicated. The revised pages are identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert I I XVII XVII 1-4 1-4 1-6 1-6 1-7 1-7 5-1 5-1 5-2 5-2 5-3 5-3 5-3a 5-3a 6-19a 6-19a 6-19b 6-19b

(4) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 342, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.

(3) Deleted Per Amendment 22, 11-20-79 (4) Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this renewed license) with less than four (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this renewed license.

(5) PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Renewed License No. DPR-70 Amendment No. 342

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS .............. .. ................................................................................. 1-1 ACTION.... ....... ........... .. .. .. ................................................................................. 1-1 AXIAL FLUX DIFFERENCE .. ................................................................................. 1-1 CHANNEL CALIBRATION . .. ................................................................................. 1-1 CHANNEL CHECK ......... ..... ................................................................................ 1-1 CHANNEL FUNCTIONAL TEST ............................................................................. 1-1 CONTAINMENT INTEGRITY ................................................................................. 1-2 CORE ALTERATION .............................................................................................. 1-2 CORE OPERATING LIMITS REPORT ................................................................... 1-2 DOSE EQUIVALENT I-131 .. ................................................................................. 1-2 DOSE EQUIVALENT XE-133 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME ......................................... 1-3 FREQUENCY NOTATION . .. ................................................................................. 1-3 FULLY WITHDRAWN .. .. .. .. ................................................................................. 1-3 GASEOUS RADWASTE TREATMENT SYSTEM .................................................. 1-3 IDENTIFIED LEAKAGE .. .. .. ................................................................................. 1-3 INSERVICE TESTING PROGRAM......................................................................... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) ............................................. 1-4 OPERABLE - OPERABILITY ................................................................................. 1-4 OPERATIONAL MODE - MODE ............................................................................. 1-4 PHYSICS TESTS ......... .. .. .. ................................................................................. 1-5 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) .............................. 1-5 PRESSURE BOUNDARY LEAKAGE ..................................................................... 1-5 PROCESS CONTROL PROGRAM (PCP) .............................................................. 1-5 PURGE-PURGING ...... .. .. .. ................................................................................. 1-5 QUADRANT POWER TILT RATIO ......................................................................... 1-5 RATED THERMAL POWER . ................................................................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME ....................................................... 1-6 REPORTABLE EVENT .. .. .. ................................................................................. 1-6 SHUTDOWN MARGIN. .. .. .. ................................................................................. 1-6 SOLIDIFICATION......... .. .. .. ................................................................................. 1-6 SOURCE CHECK ........ .. .. .. ................................................................................. 1-6 STAGGERED TEST BASIS .. ................................................................................. 1-6 THERMAL POWER ..... .. .. .. ................................................................................. 1-7 UNIDENTIFIED LEAKAGE .. ................................................................................. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM ................................................ 1-7 VENTING . ....... ........... .. .. .. ................................................................................. 1-7 SALEM - UNIT 1 I Amendment No. 342

INDEX DESIGN FEATURES

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SECTION PAGE 5.1 SITE LOCATION. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . .. . . . .. . . . .5-1 5.2 CONTAINMENT Configuration . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . . . . . . . . .. . . . . . . . 5-1 Design Pressure and Temperature . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . .. . . . . . 5-4 5.3 REACTOR CORE Fuel Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . .. . . .. . . 5-4 Control Rod Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . . . . . 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . . . . . . 5-4 5.5 METEOROLOGICAL TOWER LOCATION . . . . . . . . . .. . . . .. . . . .. . . . . . . . . .. 5-5 5.6 FUEL STORAGE Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . . .. . . . . . . 5-5 Drainage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. . . . . . . . . 5-6a Capacity . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . .. . . . .. . . . . . . 5-6a 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT . . . .. . . . .. . . . .. . . . . . . . . . . . 5-6a SALEM - UNIT 1 XVII Amendment No. 342

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

INSERVICE TESTING PROGRAM 1.15.1 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

1.16 Not Used OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT 1 1-4 Amendment No. 342

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

1.29 Not Used SOLIDIFICATION 1.30 Not Used SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to either (a) an external source of increased radioactivity, or (b) an internal source of radioactivity (keep-alive source), or (c) an equivalent electronic source check.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - Unit 1 1-6 Amendment No. 342

DEFINITIONS

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage (except Reactor Coolant Pump Seal Water Injection) which is not IDENTIFIED LEAKAGE.

1.35 Not Used VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.37 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

SALEM - UNIT 1 1-7 Amendment No. 342

5.0 DESIGN FEATURES 5.1 SITE LOCATION Salem Generating Station is located in Salem County, New Jersey along the eastern shore of the Delaware River approximately 8 miles southwest of Salem, New Jersey and 18 miles south of Wilmington, Delaware.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 210 feet.

c. Minimum thickness of concrete walls = 4.5 feet.

d. Minimum thickness of concrete roof = 3.5 feet.

e. Minimum thickness of concrete floor mat = 16 feet.

f. Nominal thickness of steel liner = 1/4 to 1/2 inch.

g. Net free volume = 2.62 x 106 cubic feet SALEM - UNIT 1 5-1 Amendment No. 342

Intentionally Left Blank SALEM - UNIT 1 5-2 Amendment No. 342

Intentionally Left Blank SALEM - UNIT 1 5-3 Amendment No. 342

Intentionally Left Blank SALEM - UNIT 1 5-3a Amendment No. 342

ADMINISTRATIVE CONTROLS following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa,

2) Seal leakage rate less than or equal to 0.01 La per hour when the gap between the door seals is pressurized to 10.0 psig.

Test frequencies and applicable extensions will be controlled by the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 will be applied to the Primary Containment Leakage Rate Testing Program.

6.8.4.g Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to the members of the public from radioactive effluents as low as reasonable achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10 CFR 20, Appendix B, Table II, Column 2,

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 92-day period would exceed a suitable fraction of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, SALEM - UNIT 1 6-19a Amendment No. 342

ADMINISTRATIVE CONTROLS

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

6.8.4.h Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of the census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.4.i Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each SALEM - UNIT 1 6-19b Amendment No. 342

PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 323 Renewed License No. DPR-75

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees), dated September 29, 2021, as supplemented by letter dated November 18, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 3

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-75 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 323, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2022.02.14 Danna 11:30:00 -05'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: February 14, 2022

ATTACHMENT TO LICENSE AMENDMENT NO. 323 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Renewed Facility Operating License No. DPR-75 with the attached revised page as indicated. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3 3 Replace the following pages of Renewed Facility Operating License No. DPR-75 with the attached revised pages as indicated. The revised pages are identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert I I XVII XVII 1-4 1-4 1-6 1-6 1-7 1-7 5-1 5-1 5-2 5-2 5-3 5-3 5-3a 5-3a 6-19a 6-19a 6-19b 6-19b

(3) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source or special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 3459 megawatts (thermal).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 323, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. DPR-75 Amendment No. 323

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS ........ .. .. .. ................................................................................. 1-1 ACTION.... ....... ........... .. .. .. ................................................................................. 1-1 AXIAL FLUX DIFFERENCE .. ................................................................................. 1-1 CHANNEL CALIBRATION . .. ................................................................................. 1-1 CHANNEL CHECK ...... .. .. .. ................................................................................. 1-1 CHANNEL FUNCTIONAL TEST ............................................................................. 1-1 CONTAINMENT INTEGRITY ................................................................................. 1-2 CORE ALTERATION ... .. .. .. ................................................................................. 1-2 CORE OPERATING LIMITS REPORT ................................................................... 1-2 DOSE EQUIVALENT I-131 .. ................................................................................. 1-2 DOSE EQUIVALENT XE-133 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME ......................................... 1-3 FREQUENCY NOTATION . .. ................................................................................. 1-3 FULLY WITHDRAWN .. .. .. .. ................................................................................. 1-3 GASEOUS RADWASTE TREATMENT SYSTEM .................................................. 1-3 IDENTIFIED LEAKAGE .. .. .. ................................................................................. 1-3 INSERVICE TESTING PROGRAM......................................................................... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) ............................................. 1-4 OPERABLE - OPERABILITY ................................................................................. 1-4 OPERATIONAL MODE - MODE ............................................................................. 1-4 PHYSICS TESTS ......... .. .. .. ................................................................................. 1-5 PRESSURE BOUNDARY LEAKAGE ..................................................................... 1-5 PROCESS CONTROL PROGRAM (PCP) .............................................................. 1-5 PURGE-PURGING ...... .. .. .. ................................................................................. 1-5 QUADRANT POWER TILT RATIO ......................................................................... 1-5 RATED THERMAL POWER . ................................................................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME ....................................................... 1-6 REPORTABLE EVENT .. .. .. ................................................................................. 1-6 SHUTDOWN MARGIN. .. .. .. ................................................................................. 1-6 SOLIDIFICATION......... .. .. .. ................................................................................. 1-6 SOURCE CHECK ........ .. .. .. ................................................................................. 1-6 STAGGERED TEST BASIS .. ................................................................................. 1-6 THERMAL POWER ..... .. .. .. ................................................................................. 1-7 UNIDENTIFIED LEAKAGE .. ................................................................................. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM ................................................ 1-7 VENTING . ....... ........... .. .. .. ................................................................................. 1-7 SALEM - UNIT 2 I Amendment No. 323

INDEX DESIGN FEATURES

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SECTION PAGE 5.1 SITE LOCATION ............................................. ......... .......... .......................... ......5-1 5.2 CONTAINMENT Configuration ............................................. ... ...................................... ...................5-1 Design Pressure and Temperature ...... ............................................. .....................5-4 5.3 REACTOR CORE Fuel Assemblies ............. ............ ............................................................ ...............5-4 Control Rod Assemblies .............. ........................... ............................ ...................5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ........... ................................ ........ ....................5-4 5.5 METEOROLOGICAL TOWER LOCATION ............ .................... ..........................5-5 5.6 FUEL STORAGE Criticality ............................... ............... .................................... ..............................5-5 Drainage ............... ........................................... ......................... .............................5-5b Capacity ................. ........................................... ............................. ........................5-5b 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...... .................. ...........................5-5b SALEM - UNIT 2 XVII Amendment No 323

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

INSERVICE TESTING PROGRAM 1.15.1 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

1.16 Not Used OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s), and when all necessary attendant instrumentation, controls, normal or emergency electrical power source, cooling and seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT 2 1-4 Amendment No. 323

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

1.29 Not Used SOLIDIFICATION 1.30 Not Used SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to either (a) an external source of increased radioactivity, or (b) an internal source of radioactivity (keep-alive source), or (c) an equivalent electronic source check.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 2 1-6 Amendment No. 323

DEFINITIONS

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage (except Reactor Coolant Pump Seal Water Injection) which is not IDENTIFIED LEAKAGE.

1.35 Not Used VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.37 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

SALEM - UNIT 2 1-7 Amendment No. 323

5.0 DESIGN FEATURES 5.1 SITE LOCATION Salem Generating Station is located in Salem County, New Jersey along the eastern shore of the Delaware River approximately 8 miles southwest of Salem, New Jersey and 18 miles south of Wilmington, Delaware.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 210 feet.

c. Minimum thickness of concrete walls = 4.5 feet.

d. Minimum thickness of concrete roof = 3.5 feet.

e. Minimum thickness of concrete floor mat = 16 feet.

f. Nominal thickness of steel liner = 1/4 to 1/2 inch.

g. Net free volume = 2.62 x 106 cubic feet SALEM - UNIT 2 5-1 Amendment No. 323

Intentionally Left Blank SALEM - UNIT 2 5-2 Amendment No. 323

Intentionally Left Blank SALEM - UNIT 2 5-3 Amendment No. 323

Intentionally Left Blank SALEM - UNIT 2 5-3a Amendment No. 323

ADMINISTRATIVE CONTROLS acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests;