ML21110A052: Difference between revisions

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3.2.9    Deletion of E Bar Determination Surveillance The combination of the current Salem SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) and Item 3 of Table 4.4-4, used for  determination, is equivalent to the initial ISTS, SR 3.4.16.3, the surveillance for the determination of . As with the TSTF-490 change, this surveillance is deleted. This is acceptable since TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) on RCS specific activity supports the dose analyses for DBAs in which the whole-body dose is primarily dependent on the noble gas concentration, not the non-gaseous activity currently captured in the  definition. With the elimination of the limit for RCS gross specific activity and the addition of the new LCO limit for noble gas specific activity, this SR to determine  is no longer required.
3.2.9    Deletion of E Bar Determination Surveillance The combination of the current Salem SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) and Item 3 of Table 4.4-4, used for  determination, is equivalent to the initial ISTS, SR 3.4.16.3, the surveillance for the determination of . As with the TSTF-490 change, this surveillance is deleted. This is acceptable since TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) on RCS specific activity supports the dose analyses for DBAs in which the whole-body dose is primarily dependent on the noble gas concentration, not the non-gaseous activity currently captured in the  definition. With the elimination of the limit for RCS gross specific activity and the addition of the new LCO limit for noble gas specific activity, this SR to determine  is no longer required.
With the relocation of Salem TS Table 4.4-4 Items 1, 2, 3, and 4b to SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) and SR 4.4.8.2 (Salem Unit No. 1) / SR 4.4.9.2 (Salem Unit No. 2), and the relocation of Salem Table 4.4-4 Item 4a to Action a.1, SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2), and Table 4.4-4 are no longer needed and the NRC staff finds their deletion acceptable.
With the relocation of Salem TS Table 4.4-4 Items 1, 2, 3, and 4b to SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) and SR 4.4.8.2 (Salem Unit No. 1) / SR 4.4.9.2 (Salem Unit No. 2), and the relocation of Salem Table 4.4-4 Item 4a to Action a.1, SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2), and Table 4.4-4 are no longer needed and the NRC staff finds their deletion acceptable.
3.3      Technical Evaluation Conclusion The NRC staff reviewed the licensees changes to revise the Salem TSs with changes equivalent to those found in TSTF-490. This includes replacing the current specific activity of the primary coolant limits with limits on primary coolant DEI and DEX specific activity. The NRC staff reviewed the proposed changes to (1) delete the primary coolant gross activity limit, its associated conditions, required actions, and SRs; (2) delete the determination of ; and (3) add a new DEX limit, its associated conditions, required actions, and SRs, and has concluded that the changes are consistent with the methodology used to analyze the radiological consequences of the SGTR and MSLB accidents.
 
===3.3      Technical Evaluation Conclusion===
The NRC staff reviewed the licensees changes to revise the Salem TSs with changes equivalent to those found in TSTF-490. This includes replacing the current specific activity of the primary coolant limits with limits on primary coolant DEI and DEX specific activity. The NRC staff reviewed the proposed changes to (1) delete the primary coolant gross activity limit, its associated conditions, required actions, and SRs; (2) delete the determination of ; and (3) add a new DEX limit, its associated conditions, required actions, and SRs, and has concluded that the changes are consistent with the methodology used to analyze the radiological consequences of the SGTR and MSLB accidents.
The NRC staff reviewed the proposed changes and determined, based on the preceding discussion, that the TSs, as revised by the proposed changes, meet the standards for TSs in 10 CFR 50.36(b) and 10 CFR 50.36(c). Further, based on the preceding discussion, the NRC staff concludes that the proposed SRs assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the TS were reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0, and were found to be acceptable. The NRC staff concludes that the TS, as
The NRC staff reviewed the proposed changes and determined, based on the preceding discussion, that the TSs, as revised by the proposed changes, meet the standards for TSs in 10 CFR 50.36(b) and 10 CFR 50.36(c). Further, based on the preceding discussion, the NRC staff concludes that the proposed SRs assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the TS were reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0, and were found to be acceptable. The NRC staff concludes that the TS, as


Latest revision as of 00:01, 23 May 2023

Issuance of Amendment Nos. 337 and 318, Revise Technical Specifications to Adopt TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec
ML21110A052
Person / Time
Site: Salem  PSEG icon.png
Issue date: 07/19/2021
From: James Kim
Plant Licensing Branch 1
To: Carr E
Public Service Enterprise Group
Kim J
References
EPID L-2020-LLA-0206
Download: ML21110A052 (37)
v • d • e


Text

July 19, 2021 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 337 AND 318 RE: REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-490, DELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY TECH SPEC (EPID L-2020-LLA-0206)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 337 and 318 to Renewed Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, respectively. These amendments consist of changes to the technical specifications (TSs) in response to your application dated September 17, 2020, as supplemented by letter dated June 9, 2021.

The amendments replace the current TS limit on the reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity would be based on a new dose equivalent xenon (Xe)-133 (DEX) definition that would replace the current E-Bar average disintegration energy definition. The proposed changes are consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.

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. 50-272 and 50-311

Enclosures:

1. Amendment No. 337 to DPR-70
2. Amendment No. 318 to DPR-75
3. Safety Evaluation cc: Listserv

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 337 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 17, 2020, as supplemented by letter dated June 9, 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. 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. 337, 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 G. James G. Danna Date: 2021.07.19 Danna 11:36:32 -04'00' 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: July 19, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 337 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 the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert I I 1-2 1-2 1-3 1-3 3/4 4-20 3/4 4-20 3/4 4-22 3/4 4-22 3/4 4-23 3/4 4-23

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. 337, 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. 337

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 MEMBER(S) OF THE PUBLIC .. ...........................................................................................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 SITE BOUNDARY ............... .. .. ...........................................................................................1-6 SOLIDIFICATION ................ .. .. ...........................................................................................1-6 SOURCE CHECK ................ .. .. ...........................................................................................1-6 STAGGERED TEST BASIS .. .. ...........................................................................................1-6 THERMAL POWER ............. .. .. ...........................................................................................1-7 UNIDENTIFIED LEAKAGE.. .. .. ...........................................................................................1-7 UNRESTRICTED AREA ...... .. .. ...........................................................................................1-7 VENTILATION EXHAUST TREATMENT SYSTEM ..............................................................1-7 VENTING ... ........ ................ .. .. ...........................................................................................1-7 SALEM - UNIT 1 Amendment No. 337

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or
b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CORE ALTERATION 1.8 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using the Thyroid Committed Dose Equivalent (CDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion.

SALEM - UNIT 1 1-2 Amendment No. 337

DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at a minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water and Soil.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 230 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No. 337





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SALEM - UNIT 1 3/4 4-22 Amendment No. 337

This page intentionally left blank SALEM - UNIT 1 3/4 4-23 Amendment No. 337

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 318 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 17, 2020, as supplemented by letter dated June 9, 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-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. 318, 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 G. James G. Danna Date: 2021.07.19 Danna 13:14:19 -04'00' 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: July 19, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 318 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 the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert I I 1-2 1-2 1-3 1-3 3/4 4-23 3/4 4-23 3/4 4-25 3/4 4-25 3/4 4-26 3/4 4-26

(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 Commission's 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. 318, 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. 318

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 MEMBER(S) OF THE PUBLIC ............................................................................... 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 SITE BOUNDARY ........ .. .. .. ................................................................................. 1-6 SOLIDIFICATION......... .. .. .. ................................................................................. 1-6 SOURCE CHECK ........ .. .. .. ................................................................................. 1-6 STAGGERED TEST BASIS .. ................................................................................. 1-6 THERMAL POWER ..... .. .. .. ................................................................................. 1-7 UNIDENTIFIED LEAKAGE .. ................................................................................. 1-7 UNRESTRICTED AREA . .. .. ................................................................................. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM ................................................ 1-7 VENTING . ....... ........... .. .. .. ................................................................................. 1-7 SALEM - UNIT 2 I Amendment No. 318

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or
b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CORE ALTERATION 1.8.1 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using the Thyroid Committed Dose Equivalent (CDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion.

SALEM - UNIT 2 1-2 Amendment No. 318

DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at a minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 230 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 2 1-3 Amendment No. 318



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SALEM - UNIT 2 3/4 4-25 Amendment No. 318

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SALEM - UNIT 2 3/4 4-26 Amendment No. 318

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 337 AND 318 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated September 17, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20266G247), as supplemented by letter dated June 9, 2021 (ADAMS Accession No. ML21160A041), PSEG Nuclear, LLC (PSEG, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for Salem Generating Station (Salem), Unit Nos. 1 and 2. The proposed changes would revise technical specification (TS) requirements relating to reactor coolant system (RCS) primary coolant activity limits. The changes are consistent with Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications (ISTS) Change Traveler, TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec (ADAMS Accession No. ML052630462). The proposed Salem changes would replace the current TS limit on primary coolant gross specific activity with a new limit on primary coolant noble gas specific activity. The noble gas specific activity limit would be based on a new Dose Equivalent Xe [xenon]-133 (DEX) definition that would replace the current [E-Bar] Average Disintegration Energy definition, consistent with TSTF-490.

For background purposes, the NRC staff notes that by letter dated September 13, 2005 (ADAMS Accession No. ML052630462), the TSTF submitted TSTF-490 for NRC staff review.

The NRC staff published a Notice of Availability on March 19, 2007 (72 FR 12838), providing notice that the NRC staff had prepared a model LAR, model safety evaluation (SE), and model proposed no significant hazards consideration (NSHC) determination, reflecting the staffs approval of TSTF-490. TSTF-490 involves changes to NUREG-1430, NUREG-1431, and NUREG-1432, Standard Technical Specifications, Section 3.4.16, RCS Specific Activity, RCS gross specific activity limits with the addition of a new limit for noble gas specific activity.

The noble gas specific activity limit would be based on a new dose equivalent Xe-133 (DEX) definition that replaces the current [E-bar] average disintegration energy definition.

Note: There are differences in the TS numbering systems for Salem Unit No. 1 and Salem Unit No. 2. For example, Limiting Condition for Operation (LCO) 3.4.8 for Salem Unit No. 1 is Enclosure 3

identical to LCO 3.4.9 for Salem Unit No. 2. In this SE, when addressing identical elements in Salem Unit No. 1 and Salem Unit No. 2 TS, the following convention will be used: (LCO) 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2).

The supplemental letter dated June 9, 2021, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed NSHC determination as published in the Federal Register on November 3, 2020 (85 FR 69656).

2.0 REGULATORY EVALUATION

The NRC staff evaluated the impact of the proposed changes as they relate to the radiological consequences of affected design-basis accidents (DBAs) that use the RCS inventory as the source term.

2.1 Functional Description The source term assumed in radiological analyses should properly be based on the activity associated with either the projected fuel damage or the maximum RCS TS values, whichever maximizes the radiological consequences. The TS limits on RCS specific activity ensure that calculated offsite doses are appropriately limited for accidents that are based on releases from the RCS without a significant amount of fuel damage.

The steam generator tube rupture (SGTR) accident and main steam line break (MSLB) accident typically do not result in fuel damage; therefore, the radiological consequence analyses are generally based on the release of primary coolant activity at maximum TS limits. For accidents that result in fuel damage, the additional dose contribution from the initial activity in the RCS is not normally evaluated and is considered to be insignificant in relation to the dose resulting from the release of fission products from the damaged fuel.

For licensees using the alternative source term (AST) in their dose consequence analyses, the NRC staff uses the regulatory guidance provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition (SRP), Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, dated July 2000 (ADAMS Accession No. ML003734190), and the methodology and assumptions stated in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792). Licensees using the AST are evaluated against the dose criteria specified in Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.67, Accident source term.

2.2 Description of the Proposed TS Changes The Salem TSs use different numbering and format than the ISTS for RCS specific activity.

This different numbering and format scheme results in a different description of the changes to the TSs and a different, but equivalent, final TS compared to those used in the TSTF-490 model SE. These differences are administrative in nature and do not affect the applicability of TSTF-490 to the Salem TSs or the associated technical justification from the model SE.

The licensee proposed the following changes to the Salem TSs:

1. The definition of - Average Disintegration Energy is removed from TS Section 1.0, Definitions, and the TS Index is updated to reflect this change.
2. A definition of Dose Equivalent XE-133 is added to TS Section 1.0 and the TS Index is updated to reflect this change.
3. The definition of DOSE EQUIVALENT I-131 is changed from:

DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (microcuries per gram) 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 thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance Report No. 11 (FGR 11), Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion.

To:

DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using the Thyroid Committed Dose Equivalent (CDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

4. Limiting Conditions for Operation (LCO) 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2), Specific Activity, is revised from:

The specific activity of the primary coolant shall be limited to:

a. 1.0 µCi/gram DOSE EQUIVALENT I-131, and
b. 100/µCi/gram.

To:

The specific activity of the primary coolant shall be limited to:

a. 1.0 µCi/gram DOSE EQUIVALENT I-131, and
b. 600 µCi/gm DOSE EQUIVALENT XE-133
5. LCO 3.4.8 (Salem Unit No.1) / LCO 3.4.9 (Salem Unit No. 2) Applicability is revised from:

MODES 1, 2, 3, 4 and 5

To:

MODES 1, 2, 3 and 4

6. The current ACTIONS for LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) are removed.
7. The following ACTIONS for LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) are proposed to state:

NOTE Specification 3.0.4.c is applicable

a. With the specific activity of the primary coolant > 1.0 µCi/gram DOSE EQUIVALENT I-131:
1. Verify DOSE EQUIVALENT I-131 60 µCi/gram once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and restore DOSE EQUIVALENT I-131 to 1.0 µCi/gram within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or
2. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With the specific activity of the primary coolant > 600 µCi/gram DOSE EQUIVALENT XE-133:
1. Restore DOSE EQUIVALENT XE-133 to 600 µCi/gram within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or
2. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
8. Current Surveillance Requirement (SR) 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) and the associated Tables, 4.4-4, are removed and the TS Index is updated to reflect this change.
9. Figure 3.4-1 is removed and the TS index is updated to reflect this change for both units.
10. New SRs 4.4.8.1 and 4.4.8.2 (Salem Unit No. 1) / SRs 4.4.9.1 and 4.4.9.2 (Salem Unit No. 2) are added:

NOTE SR 4.4.(8/9).1 is not required to be performed in MODE 4, and is not required to be performed in MODE 3 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Tavg 500°F.

4.4.8.1 Verify the specific activity of the primary coolant 600 µCi/gm DOSE EQUIVALENT XE-133 in accordance with the Surveillance Frequency Control Program.

4.4.8.2 Verify the specific activity of the primary coolant 1.0 µCi/gm DOSE EQUIVALENT I-131 in accordance with the Surveillance Frequency Control Program, and between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 15% RATED THERMAL POWER within a one hour period.

2.3 Regulatory Requirements and Guidance 2.3.1 Regulatory Requirements The NRC staffs evaluation is based upon the following regulations:

Section 50.36(a)(1) of 10 CFR, under which an applicant for an operating license must include proposed TS in its application in accordance with the requirements of 10 CFR 50.36 and [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls. . . . However, per 10 CFR 50.36(a)(1), these TS bases shall not become part of the technical specifications.

Section 50.36(b) of 10 CFR, which states that each license authorizing reactor operation will include TSs derived from the analyses and evaluation included in the safety analysis report and amendments thereto.

Section 50.36(c) of 10 CFR, which requires that TS include certain items. Per 10 CFR 50.36(c)(2)(i), Limiting conditions for operation, the TSs must include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The provision also requires that [w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the [TSs] until the condition can be met.

Section 50.36(c)(3) of 10 CFR, Surveillance requirements, under which TSs must include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Section 50.67 of 10 CFR, which states that any licensee that was initially authorized to operate prior to January 10, 1997, and who seeks to revise the current accident source term in its design basis radiological consequence analyses, must apply for a license amendment under 10 CFR 50.90, Application for amendment or license, construction permit, or early site permit.

The regulation in 10 CFR 50.67(b)(2) states that, The NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product

release, would not receive a radiation dose in excess of 0.25 Sv [Sievert]

(25 rem [roentgen Equivalent man]) total effective dose equivalent (TEDE).

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

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

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Criterion 19, Control room, which states, in part:

. . . holders of operating licenses using an alternative source term under

§ 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 [10 CFR 50.2, Definitions,] for the duration of the accident.

2.3.2 Regulatory Guidance The NRC staffs evaluation is further based upon the following RG and SRP sections:

RG 1.183, which provides guidance to licensees on performing evaluations and analyses in support of the implementation of an AST.

SRP Section 15.0.1, which provides guidance for an application for the initial implementation of an AST at operating power reactors and subsequent license amendment requests from these plants regarding the NRC staffs review of said applications.

The NRC staffs guidance for review of TSs in Chapter 16.0, Technical Specifications, of the SRP, Revision 3, dated March 2010 (ADAMS Accession No. ML100351425).

The NRC staff also considered relevant information in the Salem Updated Final Safety Analysis Report, which describes the facilitys DBAs and evaluates their radiological consequences.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.0 of this SE and TSTF-490, Revision 0. In accordance with 10 CFR 50.92, Issuance of amendment, in determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent

applicable and appropriate. In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54, Conditions of licenses.

3.1 TSTF-490 Background The primary coolant specific activity level is used in DBA analyses to determine the radiological consequences of accidents that involve the release of primary coolant activity with no substantial amount of fuel damage. For events that also include significant amounts of fuel damage, the contribution from the initial activity in the primary coolant (i.e., the level of activity prior to fuel damage) is considered insignificant and is not normally evaluated.

The maximum allowable primary coolant specific activity is governed by TSs. Due to the importance of iodine in the dose consequence analyses, the TSs specify separate dose equivalent specific activity limits for the iodine isotopes and non-iodine isotopes. The limit for iodine isotopes is specified in units of Dose Equivalent I-131 (DEI), which is the normalized quantity of I-131 that would result in the same dose consequence as the combination of the major isotopes of iodine present in the primary coolant. The TSs for DEI include both an equilibrium long-term limit as well as a higher maximum allowable short-term limit, to account for iodine spiking.

The limit for non-iodine isotopes has traditionally been based on an evaluation of the average beta and gamma disintegration energy of the total non-iodine activity in the RCS, which is referred to as . Salems TSs define as shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. The RCS non-iodine specific activity limit is then expressed as the quantity 100 divided by in units of microcuries per gram. In DBA dose consequence analyses, based on releases from the RCS with no significant fuel damage, the concentration of noble gas activity in the coolant is derived from that level associated with 1 percent fuel clad defects. Based on operating experience, depending on the isotopes used to calculate and the actual degree of fuel clad defects, the routinely calculated value of may not effectively indicate the level of noble gas activity relative to the levels used in the DBA dose consequence analyses on which the limit is based.

3.2 Technical Evaluation of TSTF-490 TS Changes 3.2.1 Deletion of the Definition of and the Addition of a New Definition for DEX The new definition for DEX is similar to the definition for DEI. The determination of DEX will be performed in a similar manner to that currently used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases krypton (Kr)-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present, which are significant in terms of contribution to whole-body dose.

Some noble gas isotopes are not included due to low concentration, short half-life, or small dose conversion factor. The calculation of DEX would use the effective dose conversion factors for air submersion from Table III.1 of the U.S. Environmental Protection Agency (EPA) Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

Using this approach, the limit on the amount of noble gas activity in the primary coolant would not fluctuate with variations in the calculated values of . If a specified noble gas nuclide is not

detected, the new definition states that it should be assumed that the nuclide is present at the minimum detectable activity. This will result in a conservative calculation of DEX.

When is determined using a design-basis approach in which it is assumed that 1.0 percent of the power is being generated by fuel rods having cladding defects, and it is also assumed that there is no removal of fission gases from the letdown flow, the value of is dominated by Xe-133. The other nuclides have relatively small contributions. However, during normal plant operation, there are typically only a small amount of fuel clad defects, and the radioactive nuclide inventory can become dominated by tritium and corrosion and/or activation products, resulting in the determination of a value of that is very different than would be calculated using the design-basis approach. Because of this difference, the accident dose analyses become disconnected from plant operation, and the LCO becomes essentially meaningless. This difference also results in a TS limit that can vary during operation as different values for are determined.

This proposed change will implement an LCO that is consistent with the whole-body radiological consequence analyses, which are sensitive to the noble gas activity in the primary coolant but not to other non-gaseous activity currently captured in the definition. TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) specifies the limit for primary coolant specific activity of the RCS as 100/ µCi/gram. The current definition includes radioisotopes that decay by the emission of both gamma and beta radiation. The current Action b of TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) would rarely, if ever, be entered for exceeding 100/ µCi/gram, since the calculated value is very high (the denominator is very low) if beta emitters such as tritium are included in the determination, as required by the definition.

The following TS 1.11 definition -AVERAGE DISINTEGRATION ENERGY is deleted:

shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

The new definition DOSE EQUIVALENT XE-133, replaces the above -AVERAGE DISINTEGRATION ENERGY definition. The added definition for DEX states:

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

The licensees proposed deletion of the above-stated definition for and addition of a new definition for DEX in TS 1.11 is acceptable from a radiological dose perspective since it will result in an LCO that more closely relates the non-iodine RCS activity limits to the dose consequence analyses that form their bases.

3.2.2 Revision of TS 3.4.8, LCO Salem TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) are modified to specify that iodine specific activity in terms of DEI shall be 1.0 µCi/gram and noble gas specific activity in terms of DEX shall be 600 µCi/gm. The NRC staff finds that this proposed change is technically equivalent to the current LCO and clarifies that the primary coolant specific activity limits are DEI and DEX and thus is acceptable.

3.2.3 TS 3.4.8 Applicability The current TS 3.4.8 (Salem Unit No. 1) / TS 3.4.9 (Salem Unit No. 2) has one set of action statements with applicability MODES 1, 2 and 3* and a second action statement with applicability MODES 1, 2, 3, 4, and 5. The first set of actions only apply with the RCS average temperature (Tavg) greater than or equal to 500 degrees Fahrenheit (°F). TS 3.4.8 / TS 3.4.9 Applicability is modified to remove the different sets of action statements based on Tavg and to remove Mode 5, such that there is now one group of action statements with applicability to Modes 1-4. The LCO needs to apply during Modes 1 through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these Modes. In Mode 5, with the RCS loops filled, the steam generators are specified as a backup means of decay heat removal by natural circulation. In this Mode, however, due to the reduced temperature of the RCS, the probability of a DBA involving the release of significant quantities of RCS inventory is greatly reduced. Therefore, monitoring of RCS specific activity is not required. In Mode 5, with the RCS loops not filled, the steam generators are not used for decay heat removal, the RCS and steam generators are depressurized, and primary-to-secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required. The proposed change to modify the LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) Applicability to include Modes 1, 2, 3, and 4 is necessary to limit the potential radiological consequences of an SGTR or MSLB that may occur during these Modes and the proposed change is, therefore, acceptable from a radiological dose perspective.

3.2.4 TS 3.4.8 Action a Revision The Salem TS 3.4.8 / TS 3.4.9 current Action statements use a different numbering and format scheme than the ISTS. Both current Salem Action as contain equivalent actions to the initial ISTS Condition A (i.e., DEI greater than 1 microcurie per gram). The language in those action statements equivalent to ISTS Condition A reads, With the specific activity of the primary coolant > 1 µCi/gram DOSE EQUIVALENT I-131. These actions are replaced with a new Action a that is equivalent to the post-TSTF-490 ISTS Condition A. The site-specific DEI limit of 1 µCi/gram is maintained in Action a, as well as repeated in new SR 4.4.8.2. These numbering and format changes have no impact from a radiological dose perspective.

The current Salem Action a that is applicable in MODES 1, 2, and 3 with Tavg > 500 °F and Action a that is applicable in MODES 1, 2, 3, 4, and 5 contain actions equivalent of the initial ISTS without TSTF-490 implemented. Under initial ISTS Required Action A.1, licensees must

[v]erify Dose Equivalent I-131 within the acceptable region of Figure 3.4.16-1, with a Completion Time (CT) of once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The Salem current Action a requires a shutdown for exceeding the limit line shown on Figure 3.4-1, Dose Equivalent I-131 Primary Coolant Specific Activity Limit Versus Percent of Rated Thermal Power with the Primary Coolant Specific Activity > 1 µCi/gram Dose Equivalent I-131. Salem Figure 3.4-1 is equivalent to the initial ISTS Figure 3.4.16-1. The Salem Action a applicable in MODES 1, 2, 3, 4, and 5 requires the performance of the sampling and analysis requirements of Item 4.a) of Table 4.4-4, until the

specific activity of the primary coolant is restored to within its limits. Salem Table 4.4-4, PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM, Item 4.a) requires, in part, the isotopic analysis for iodine once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds 1 µCi/gram DEI. When both Action as are combined, they are equivalent to ISTS Required Action A.1. The new Action a.1 removes the reference to Figure 3.4-1 and inserts a limit of less than or equal to the site-specific DEI spiking limit, which is 60 microcuries per gram.

The TSTF-490 change to the ISTS requires licensees to Verify DOSE EQUIVALENT I-131

[60] µCi/gram. These two changes are, therefore, equivalent.

The radiological dose consequence analyses for the SGTR and MSLB accidents take into account the pre-accident iodine spike and do not consider the elevated RCS iodine specific activities permitted by Figure 3.4-1 for operation at power levels below 80 percent rated thermal power. Instead, the pre-accident iodine spike analyses assume a DEI concentration 60 times higher than the corresponding long-term equilibrium value, which corresponds to the specific activity limit associated with 100 percent rated thermal power operation. It is acceptable that TS 3.4.8 Action a.1 (Salem Unit No. 1) / TS 3.4.9 Action a.1 (Salem Unit No. 2) is based on the short-term, site-specific DEI spiking limit to be consistent with the assumptions contained in the radiological consequence analyses.

The current Salem Action a states, With the specific activity of the primary coolant > 1 µCi/gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, . . . .

This action is equivalent to the initial ISTS Required Action A.2 (i.e., Restore Dose Equivalent I-131 to within limit with a CT of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />). Salem Action a is replaced with a new Action a.1 that contains requirements equivalent to the ISTS change in TSTF-490. This numbering and format change has no impact from a radiological dose perspective.

The Salem proposed Action a.2. which states Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. aligns with ISTS, as modified by TSTF-490.

Condition C of the ISTS, as modified by TSTF-490, requires licensees to place the plant in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after DEI exceeds [60] µCi/gram. The allowed CTs are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

3.2.5 TS 3.4.8 Action b Revision to Include Action for DEX Limit Salem TS 3.4.8 Action b is modified to provide an Action for when DEX is not 600 µCi/ gram, and to remove the limit associated with gross activity of the reactor coolant (). This change is made to be consistent with the change to TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2), which requires the DEX specific activity to be 1.0 µCi/ gram, as discussed above in Section 3.2.2. This change is equivalent to the change to Condition B in the ISTS implemented by TSTF-490. The DEX limit is site-specific, and the numerical value of 600 µCi/

gram is contained in new SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2). The site-specific limit of 600 µCi/ gram DEX is established based on the maximum accident analysis RCS activity corresponding to 1 percent fuel clad defects with sufficient margin to accommodate the exclusion of those isotopes based on low concentration, short half-life, or small dose conversion factors. The primary purpose of TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) on RCS specific activity and its associated conditions is to support the dose analyses for DBAs. The whole-body dose is primarily dependent on the noble gas activity, not the non-gaseous activity currently captured in the definition.

The CT for new TS 3.4.8 Action b.1 requires restoration of DEX to 600 µCi/gram in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

This is consistent with the proposed CT for TS 3.4.8 Action a.1 for DEI. The radiological consequences for the SGTR and the MSLB accidents demonstrate that the calculated thyroid doses are generally a greater percentage of the applicable acceptance criteria than the calculated whole-body doses. It then follows that the CT for noble gas activity being out of specification in new SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) should be at least as great as the CT for iodine specific activity being out of specification in revised Action a.1. Therefore, the CT of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for new Action b.1 is acceptable from a radiological dose perspective because it is expected that if there were a xenon spike in the normal operation primary coolant, iodine concentration would be restored within this time period. In addition, there is a low probability of an SGTR or an MSLB occurring during this time period.

Similar to proposed Action a.2, the Salem proposed Action b.2 which states Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. aligns with ISTS, as modified by TSTF-490. Condition C of ISTS, as modified by TSTF-490, requires licensees to place the plant in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after DEX exceeds the limit and cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

3.2.6 Reformat of TS 3.4.8 Action c The Salem TS 3.4.8 Action c, which states, Specification 3.0.4.c is applicable, is equivalent to the ISTS Note in the Required Action for Condition A without TSTF-490 implemented and the TSTF-490 proposed Note to be added to the Required Action for Condition B in the ISTS. This Action currently allows entry into a MODE or other specified condition in the LCO Applicability when LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) is not being met. The text of Salem TS 3.4.8 current Action c is reformatted as a NOTE above new Actions a and b. This is to reformat the Actions for Salem TS 3.4.8 to be closer to the equivalent in ISTS. The proposed Note would allow entry into the applicable MODES from Mode 4 to Mode 1 (power operation) while the DEX limit is exceeded and the DEX is being restored to an activity level within its limit.

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

3.2.7 SR 4.4.8.1/4.4.9.1 DEX Surveillance The combination of the current LCO 3.4.8.b (Salem Unit No. 1) / LCO 3.4.9.b (Salem Unit No. 2); SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2); and Item 1 of Table 4.4-4, which is the SR for Gross Activity Determination, are equivalent to the initial ISTS SR 3.4.16.1, the surveillance for RCS gross specific activity. Equivalent to the TSTF-490 change, these are replaced by a new SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) with a requirement to verify that the primary coolant DEX specific activity is 600 µCi/gm, which is the site-specific limit for Salem. This change provides a surveillance for the new LCO limit added to TS 3.4.8 (Salem Unit No. 1) / TS 3.4.9 (Salem Unit No. 2) for DEX. The revised SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) surveillances requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant in accordance with the surveillance frequency control program, which is the same frequency required under the current SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) for RCS gross non-iodine specific activity. This measurement is the sum of the degassed gamma

activities and the gaseous gamma activities in the reactor coolant sample taken to perform the revised SRs. This surveillance provides an indication of any increase in the noble gas specific activity. The results of the surveillance for DEX allow proper remedial actions to be taken before reaching the LCO limit under normal operating conditions; therefore, the deletion of SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) regarding gross radioactivity determination and the addition of SR 4.4.8.1 / SR 4.4.9.1 are acceptable.

The licensee proposed that the surveillances be performed in MODES 1, 2, 3 and 4 to ensure consistency with the LCO Applicability. This is contrary to the TSTF-490 requirement of SR 3.4.16.1 only being performed in Mode 1. In the September 17, 2020, LAR, the licensee stated the deviation from TSTF-490 is more conservative than the TSTF. The licensee also stated that Salems plant configuration supports performing the surveillances in MODES 1, 2, 3 and 4. The NRC staff determined that the deviation is acceptable because it would be a more conservative requirement than is provided in the ISTS as modified by TSTF-490.

In the June 9, 2021, supplement to the LAR, the licensee requested to add a NOTE above SR 4.4.8.1 / SR 4.4.9.1. The NOTE states, SR 4.4.(8/9).1 is not required to be performed in MODE 4, and is not required to be performed in MODE 3 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Tavg 500°F.

The licensee provided the following explanation for the need for the NOTE:

Based on further assessment of this testing, it was identified that a representative noble gas sample to assess DEX may not be obtainable at the reduced RCS pressure and temperature conditions during operations in Mode 3 with Tavg less than 500°F and in Mode 4.

Based on this technical limitation for performance of the DEX SR and the limited time the reactor operates in this Mode, PSEG is proposing the following additional change to the TS to address this issue The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance in MODE 3 provides a reasonable time in which to complete the DEX surveillance after T avg is greater than or equal to 500°F and RCS conditions permit a representative sample to be obtained. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance is acceptable based on the low likelihood of exceeding the DEX limit during this brief period in a reactor startup.

The NRC staff reviewed the licensees justification and evaluation of the NOTE proposed in its June 9, 2021, supplement. The NRC staff further evaluated the impact of the NOTE. The NRC staff determined that the NOTE modifies the required performance of the SR and is properly construed to be part of the specified frequency. The impact of the NOTE is that for the situation where the SR frequency is exceeded while Tavg is < 500 °F, this NOTE allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Tavg 500 °F to perform the SR. The SR would still be considered to be performed within the "specified Frequency." Therefore, if the SR were not performed within the specified frequency (plus the extension allowed by SR 4.0.2) interval, but operation was

< 500 °F, the requirement to perform the SR would not be violated, and there would not be a failure to meet the LCO. Once the unit reaches Tavg 500 °F, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would be allowed for completing the SR. If the SR were not performed within this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval after Tavg 500 °F (plus the extension allowed by SR 4.0.2), there would then be a failure to perform a SR within the specified Frequency, and the provisions of SR 4.0.3 would apply.

The NRC staff determined the NOTE is appropriate because there is no value in performing the SR when plant conditions do not exist to allow collecting a representative RCS sample. The

NRC staff determined that the NOTE is acceptable because it acceptably modifies the required performance of the SR while maintaining all other SR performance rules.

3.2.8 SR 4.4.8.2/4.4.9.2 DEI Surveillance The combination of the current Salem TS LCO 3.4.8.a (Salem Unit No. 1) / TS LCO 3.4.9.a (Salem Unit No. 2), SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2), and Items 2 and 4.b of Table 4.4-4 are equivalent to initial ISTS SR 3.4.16.2, the surveillance for DEI specific activity. These SRs are revised with a new SR 4.4.8.2 (Salem Unit No. 1) / SR 4.4.9.2 (Salem Unit No. 2), with these revised SRs replacing both Item 2 of Table 4.4-4 and Item 4.b of Table 4.4-4. This is a format change to more closely resemble the corresponding SR in the ISTS and is administrative in nature and is therefore acceptable.

3.2.9 Deletion of E Bar Determination Surveillance The combination of the current Salem SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2) and Item 3 of Table 4.4-4, used for determination, is equivalent to the initial ISTS, SR 3.4.16.3, the surveillance for the determination of . As with the TSTF-490 change, this surveillance is deleted. This is acceptable since TS LCO 3.4.8 (Salem Unit No. 1) / LCO 3.4.9 (Salem Unit No. 2) on RCS specific activity supports the dose analyses for DBAs in which the whole-body dose is primarily dependent on the noble gas concentration, not the non-gaseous activity currently captured in the definition. With the elimination of the limit for RCS gross specific activity and the addition of the new LCO limit for noble gas specific activity, this SR to determine is no longer required.

With the relocation of Salem TS Table 4.4-4 Items 1, 2, 3, and 4b to SR 4.4.8.1 (Salem Unit No. 1) / SR 4.4.9.1 (Salem Unit No. 2) and SR 4.4.8.2 (Salem Unit No. 1) / SR 4.4.9.2 (Salem Unit No. 2), and the relocation of Salem Table 4.4-4 Item 4a to Action a.1, SR 4.4.8 (Salem Unit No. 1) / SR 4.4.9 (Salem Unit No. 2), and Table 4.4-4 are no longer needed and the NRC staff finds their deletion acceptable.

3.3 Technical Evaluation Conclusion

The NRC staff reviewed the licensees changes to revise the Salem TSs with changes equivalent to those found in TSTF-490. This includes replacing the current specific activity of the primary coolant limits with limits on primary coolant DEI and DEX specific activity. The NRC staff reviewed the proposed changes to (1) delete the primary coolant gross activity limit, its associated conditions, required actions, and SRs; (2) delete the determination of ; and (3) add a new DEX limit, its associated conditions, required actions, and SRs, and has concluded that the changes are consistent with the methodology used to analyze the radiological consequences of the SGTR and MSLB accidents.

The NRC staff reviewed the proposed changes and determined, based on the preceding discussion, that the TSs, as revised by the proposed changes, meet the standards for TSs in 10 CFR 50.36(b) and 10 CFR 50.36(c). Further, based on the preceding discussion, the NRC staff concludes that the proposed SRs assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the TS were reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0, and were found to be acceptable. The NRC staff concludes that the TS, as

amended by the proposed changes, meet the requirements stated in 10 CFR 50.36 and are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendments on December 23, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change SRs.

The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve NSHC, published in the Federal Register on November 3, 2020 (85 FR 69656), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: S. Meighan M. Hamm Date: July 19, 2021

ML21110A052 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC NAME JKim JBurkhardt VCusumano DATE 04/21/2021 04/21/2021 04/14/2021 OFFICE OGC - NLO NRR/DRA/ARCB/BC NRR/DORL/LPL2/LAiT NAME 05/6/2021; 06/21/2021 KHsueh KEntz DATE STurk 04/26/2021 06/24/2021 OFFICE NRR/DORL/LPL1/LA NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME PBlechman JDanna JKim DATE 06/30/2021 07/15/2021 07/19/2021

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