Text

License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition, Part 3 of 5
	ML20303A326
Person / Time
Site:	Byron
Issue date:	10/29/2020
From:	; Exelon Generation Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML20034A147	List: ML20303A326;
References
	RS-20-129
	Download: ML20303A326 (60)
	v • d • e

{{#Wiki_filter:ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.7 - Plant Systems (Current Title) TS Section 3.7 - Facility Systems (Proposed Title) TS 3.7.12, "Nonaccessible Area Exhaust Filter Plenum Ventilation System," requires three Nonaccessible Area Exhaust Filter Plenum Ventilation System trains to be operable, with two trains aligned for operation and one train aligned in standby. TS 3.7.12 is applicable in MODES 1, 2, 3, and 4 . The Nonaccessible Area Exhaust Filter Plenum Ventilation System filters air from the area of the active ECCS components during the recirculation phase of a LOCA. The Nonaccessible Area Exhaust Filter Plenum Ventilation System , in conjunction with other normally operating systems , also provides environmental control of temperature in the ECCS pump room area and the lower reaches of the auxiliary building . TS 3.7.13, "Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System ," requires two FHB Ventilation system trains be operable. TS 3. 7 .13 is applicable during movement of recently irradiated fuel assemblies in the fuel building and during movement of recently irradiated fuel assemblies in containment with the equipment hatch not intact. During movement of recently irradiated fuel in the FHB , the FHB Ventilation System is required to be operable to alleviate the consequences of an FHA. In addition , the revised FHA calculation does not credit operability of the FHB Ventilation System . The results of the analysis demonstrate that the dose consequences for the main control room remain below the acceptance criteria , without relying on active components remaining functional for accident mitigation during and following the event. Proposed License Conditions 2.C.(25) and 2.C.(14) for Byron , Unit 1 and Unit 2, respectively , wil l prohibit movement of irradiated fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown. TS 3.7.14 , "Spent Fuel Pool Water Level ," is applicable to the safe storage and handling of irradiated fuel in the SFP and therefore is still applicable in the permanently defueled condition. TS 3. 7 .14 is being proposed for inclusion into the POTS , with the proposed changes described below. TS 3.7.15, "Spent Fuel Pool Boron Concentration," is applicable to the safe storage and handling of irradiated fuel in the SFP and therefore is still applicable in the permanently defueled condition. TS 3. 7 .15 is being proposed for inclusion into the POTS , with the proposed changes described below. TS 3.7.16, "Spent Fuel Assembly Storage ," is applicable to the safe storage and handling of irradiated fuel in the SFP and therefore is still applicable in the permanently defueled condition . TS 3.7 .16 is being proposed for inclusion into the POTS , with the proposed changes described below. Current TS 3.7.14 Current TS 3.7.14 ACTIONS ACTIONS

N 0 TE--------------------------------

ATC::

LCO 3.0 .3 is not applicable. bGG J.G.J is Aet a13131isa91e. Current TS 3. 7 .15 Current TS 3.7 .15 ACTIONS ACTIONS

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ATC::

LCO 3.0.3 is not applicable . bGG J.Q.J is Ret a13131isa91e. Current TS 3. 7 .16 Current TS 3.7 .16 ACTIONS ACTIONS

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ATC::

LCO 3.0.3 is not applicable. bGG d.G.d is Ret a1313lisaele. Page 57 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.7 - Plant Systems (Current Title) TS Section 3.7 - Facility Systems (Proposed Title) Summary In the Section Title, the reference to the term "PLANT" is replaced with the term "FACILITY" because the term "plant" generally refers to the reactor, which can no longer be operated , whereas the term "facility" refers to the overall site. TS 3.7.14, TS 3.7.15, and TS 3.7.16 will continue to remain applicable with the reactor permanently defueled . As such , they are being retained and revised, as necessary, to reflect a permanently defueled condition. The Notes in the Actions of TS 3.7.14, TS 3.7 .15, and TS 3.7 .16 are proposed for deletion because LCO 3.0.3 is not being retained in the POTS as discussed in the changes proposed for TS Section 3.0. Attachment 3 provides the existing Byron TS pages, marked up to show the proposed changes for TS 3.7.14 , TS 3.7.15, and TS 3. 7 .16. Proposed changes to the TS Bases addressing the proposed changes to the relevant TS are provided for information in Attachment 4. The content of TS 3.7.1 through TS 3.7 .13 are being proposed for deletion in their entirety and are not included in Attachment 3. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Byron , the 10 CFR 50 licenses will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactor vessels in accordance with 10 CFR 50.82(a)(2). As a result, these TS will not apply with the reactor defueled. The applicable Bases and surveillance sections will also be deleted to reflect this change . TS Section 3.8 - Electrical Power Systems TS Being Deleted TS Being Retained 3.8.1 - AC Sources-Operating 3.8.2 - AC Sources-Shutdown 3.8.3 - Diesel Fuel Oil 3.8.4 - DC Sources-Operatinq 3.8.5 - DC Sources-Shutdown 3.8.6 - Battery Parameters 3.8.7 - lnverters-Ooeratinq 3.8.8 - Inverters-Shutdown 3.8.9 - Distribution Svstems-Operatinq 3.8.10 - Distribution Systems-Shutdown Basis The existing TS Section 3.8, "Electrical Power Systems," contains LCOs that provide for appropriate functional capability of plant electrical equipment required for safe operation of the facility, including the plant being a defueled condition . All TS in Section 3.8 are being proposed for deletion, as identified in the table above . TS 3.8.1, "AC Sources-Operating," specifies the AC electrical power sources that shall be operable in MODES 1, 2, 3, and 4. Operability of these systems ensure that acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients and adequate core cooling is provided and containment operability and other vital functions are maintained in the event of a postulated OBA TS 3.8.2, "AC Sources-Shutdown ," identifies the AC electrical power sources that shall be operable in MODES 5 and 6 and during movement of irradiated fuel assemblies. The operability of the minimum AC sources during MODES 5 and 6 and during movement of fuel assemblies ensures that:

a. The unit can be maintained in the shutdown or refueling condition for extended periods ;

Page 58 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.8 - Electrical Power Systems

b. Sufficient instrumentation and control capability are available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as a fuel handling accident.

TS 3.8.2 is applicable in MODES 5 and 6, and during movement of irradiated fuel assemblies. The FHA in the Fuel Handling Building is a remaining postulated design basis accident event that could potentially occur at a permanently defueled facility . However, the FHA in the Fuel Handling Building does not credit any AC sources for mitigation of the accident. There are no active systems credited as part of the initial conditions of the analysis or as part of the primary success path for mitigation of the FHA in the Fuel Handling Building with Byron in the permanently shutdown and defueled condition . In addition, the mode of applicability associated with moving recently irradiated fuel is no longer applicable. Therefore, TS 3.8.2 will no longer be needed for assuring the appropriate functional capability of the AC sources for safe operation of the facility when moving recently irradiated fuel assemblies and is proposed for deletion. Proposed License Conditions 2.C.(25) and 2.C .(14) for Byron, Unit 1 and Unit 2, respectively, will prohibit movement of irradiated fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown . TS 3.8.3, "Diesel Fuel Oil ," ensures specific limits are met for the stored diesel fuel oil for each required diesel generator (DG). For proper operation of the standby DGs , it is necessary to ensure the proper quality of the fuel oil. Stored diesel fuel oil is required to have sufficient supply for 7 days of post-accident load operation . It is also required to meet specific standards for quality. This requirement, in conjunction with an ability to obtain replacement supplies within 7 days, supports the availability of DGs required to shutdown the reactor and to maintain it in a safe condition for an AOO or a postulated DBA with loss of offsite power. The AC sources (LCO 3.8.1 and LCO 3.8.2) are required to ensure the availability of the required power to shutdown the reactor and maintain it in a safe shutdown condition after an AOO or a postulated OBA. Since the stored diesel fuel oil supports LCO 3.8.1 and LCO 3.8.2 , stored diesel fuel oil is required to be within limits when the associated DG is required to be operable. TS 3.8.3 is required to support the DG requirements of TS 3.8.1 and TS 3.8.2. With the proposed deletion of TS 3.8.1 and TS 3.8.2 , the requirements of TS 3.8.3 are no longer applicable and are propose for deletion . Since TS 3.8.3 exists solely to support the DG requirements of TS 3.8 .1 and TS 3.8.2, the elimination of the need for DGs also obviates the need for the support systems. Therefore , TS 3.8.3 is proposed for deletion. Stored diesel fuel oil is required to have sufficient supply for 7 days of post accident load operation . It is also required to meet specific standards for quality. Each DG is provided with fuel oil capacity sufficient to operate that diesel for a period of 7 days while the DG is supplying the post loss of coolant accident load demand discussed in the UFSAR. TS 3.8.3 is applicable when associated DG is required to be operable. As discussed above , the requirement for DGs is being deleted from the TS because there are no DBAs or transients analyzed in UFSAR Chapter 15 that rely on the DGs for mitigation. Since TS 3.8.3 exists solely to support the DG requirements of TS 3.8.1 and TS 3.8.2, the elimination of the need for DGs also obviates the need for their support systems. As such , TS 3.8.3 may be deleted. Based on the above, the proposed deletion of TS for diesel fuel oil is acceptable . TS 3.8.4, "DC Sources Operating ," specifies requirements to ensure that the DC electrical power subsystems (with each subsystem consisting of one 125 VDC battery, the associated battery charger for each battery, and all the associated control equipment and interconnecting cabling) are required to be operable to ensure the availability of the required power to shutdown the reactor and maintain it in a safe condition after an AOO or a postulated OBA. The station DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment and AC instrument bus power (via inverters). As required by 10 CFR 50 , Appendix A, GDC 17, the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. TS 3.8.4 is applicable in MODES 1, 2, 3, and 4. Page 59 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.8 - Electrical Power Systems TS 3.8.5, "DC Sources-Shutdown ," requires one DC electric power subsystem to be operable. The operability of the minimum DC electrical power sources during MODES 5 and 6 and during movement of irradiated fuel assemblies ensures that:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown , such as a fuel handling accident.

TS 3.8.5 is applicable in MODES 5 and 6, and during movement of irradiated fuel assemblies. The FHA in the Fuel Handling Building is a remaining postulated design basis accident event that could potentially occur at a permanently defueled facility. However, the FHA in the Fuel Handling Building does not credit any DC sources for mitigation of the accident. DC sources are therefore not required during movement of irradiated fuel assemblies in the SFP for mitigation of a potential FHA. There are no active systems credited as part of the initial conditions of the analysis or as part of the primary success path for mitigation of the FHA in the Fuel Handling Building with Byron in the permanently shutdown and defueled condition . In addition , the mode of applicability associated with moving recently irradiated fuel is no longer applicable. Therefore, TS 3.8.5 will no longer be needed for assuring the appropriate functional capability of the DC sources for safe operation of the facility when moving recently irradiated fuel assemblies and is proposed for deletion. Proposed License Conditions 2.C.(25) and 2.C.(14) for Byron, Unit 1 and Unit 2, respectively, will prohibit movement of irradiated fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown . TS 3.8.6, "Battery Parameters ," specifies the battery parameters for the Division 11 (21) and Division 12(22) batteries to be within limits. TS 3.8.6 is applicable when associated DG electrical power subsystems are required to be operable. This specification delineates the limits on battery float current as well as electrolyte temperature, level, and float voltage for the DC power subsystem source batteries. Additional preventative maintenance, testing, and monitoring performed in accordance with the Battery Monitoring and Maintenance Program is conducted as specified in Specification 5.5.17 . Battery parameters must remain within acceptable limits to ensure availability of the required DC power to shutdown the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated OBA. Battery parameters are required solely for the support of the associated DC electrical power subsystems (TS 3.8.4 and TS 3.8.5) . Therefore , battery parameter limits are only required when the DC electrical power subsystems are required to be operable. As TS 3.8.4 are proposed for deletion, TS 3.8.6 is also proposed for deletion . TS 3.8.7, "Inverters-Operating ," requires four instrument bus inverters to be operable. The inverters are the preferred source of power for the AC instrument buses because of the stability and reliability they provide. Each of the four AC instrument buses (2 per division) is normally supplied AC electrical power by a dedicated inverter. The inverters can be powered from an AC source/rectifier or from an associated 125 VDC battery. The battery provides an uninterruptible power source for the instrumentation and controls for the RPS and the ESFAS. TS 3.8.7 is applicable in MODES 1, 2, 3, and 4. TS 3.8.8, "Inverters-Shutdown ," requires two inverters to be operable . The DC to AC inverters are designed to provide the required capacity, capability, redundancy, and reliability to ensure the availability of necessary power to the Reactor Protective System (RPS) and ESFAS instrumentation and controls so that the fuel, RCS , and containment design limits are not exceeded. The operability of the inverter to each required AC instrument bus during MODES 5 and 6 ensures that:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status ; and Page 60 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.8 - Electrical Power Systems

c. Adequate power is available to mitigate events postulated during shutdown, such as a fuel handling accident.

TS 3.8.8 is applicable in MODES 5 and 6, and during movement of irradiated fuel assemblies. The FHA in the Fuel Handling Building is a remaining postulated design basis accident event that could potentially occur at a permanently defueled facility. However, the inverters are not required during movement of irradiated fuel assemblies in the SFP for mitigation of a potential FHA. There are no active systems credited as part of the initial conditions of the analysis or as part of the primary success path for mitigation of the FHA in the Fuel Handling Building with Byron in the permanently shutdown and defueled condition . In addition, the mode of applicability associated with moving recently irradiated fuel is no longer applicable. Therefore, TS 3.8.8 will no longer be needed for assuring the operability of the inverters for safe operation of the facility when moving recently irradiated fuel assemblies and is proposed for deletion. Proposed License Conditions 2.C .(25) and 2.C.(14) for Byron, Unit 1 and Unit 2, respectively, will prohibit movement of irradiated fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown. TS 3.8.9, "Distribution Systems-Operating," specifies the required portions of the AC, DC, and AC instrument bus electrical power distribution subsystems to be operable for the applicable unit. TS 3.8.9 is applicable in MODES 1, 2, 3, and 4. The required power distribution subsystems ensure the availability of AC, DC, and AC instrument bus electrical power for the systems required to shutdown the reactor and maintain it in a safe condition after an AOO or a postulated OBA. The AC, DC, and AC instrument bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded. Maintaining the Division 1 and Division 2 AC, DC, and AC instrument bus electrical power distribution subsystems operable ensures that the redundancy incorporated into the design of ESF is not defeated. TS 3.8.10, "Distribution Systems-Shutdown," specifies the required portions of the AC, DC, and AC instrument bus electrical power distribution subsystems to be operable to support equipment required to be operable for the applicable unit. The AC, DC, and AC instrument bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded. The operability of the minimum AC, DC, and AC instrument bus electrical power distribution subsystems during MODES 5 and 6, and during movement of irradiated fuel assemblies ensures that:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as a fuel handling accident.

TS 3.8.10 is applicable in MODES 5 and 6, and during movement of irradiated fuel assemblies. The FHA in the Fuel Handling Building is a remaining postulated design basis accident event that could potentially occur at a permanently defueled facility . However, an FHA does not rely on electrical distribution systems for accident mitigation. Therefore, the electrical distribution systems are not required during movement of irradiated fuel assemblies in the SFP for mitigation of a potential FHA. There are no active systems credited as part of the initial conditions of the analysis or as part of the primary success path for mitigation of the FHA in the Fuel Handling Building with Byron in the permanently shutdown and defueled condition . In addition, the mode of applicability associated with moving recently irradiated fuel is no longer applicable. Therefore, TS 3.8.10 will no longer be needed for assuring the operability of the electrical distribution systems for safe operation of the facility when moving recently irradiated fuel assemblies and is proposed for deletion. Proposed License Conditions 2.C.(25) and 2.C.(14) for Byron, Unit 1 and Unit 2, respectively, will prohibit movement of irradiated Page 61 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.8 - Electrical Power Systems fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown . Summary The content of TS Section 3.8 is being proposed for deletion in its entirety. TS 3.8.1, TS 3.8.4, TS 3.8.7, and TS 3.8.9 will not apply with the reactor defueled. After the certifications required by 10 CFR 50 .82(a)(1) are submitted for Byron, the 10 CFR 50 licenses will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactor vessels in accordance with 10 CFR 50.82(a)(2). The specific summaries for TS 3.8.2, TS 3.8.3, TS 3.8.5, TS 3.8.6, TS 3.8.8, and TS 3.8.10 are provided in the table above. As a result, TS Section 3.8 will not apply and is proposed for deletion . With TS Section 3.8 deleted in its entirety, the appl icable Bases and surveillance section will also be deleted to reflect this change. TS Section 3.9 - Refueling Operations TS Being Deleted TS Being Retained 3.9.1 - Boron Concentration 3.9.2 - Unborated Water Source Isolation Valves 3.9.3 - Nuclear Instrumentation 3.9.4 - Containment Penetrations 3.9.5 - Residual Heat Removal (RHR) and Coolant Circulation-Hiqh Water Level 3.9.6 - Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level 3.9.7 - Refuelinq Cavity Water Level Basis The existing TS Section 3.9, "Refueling Operations," contains LCOs that provide for appropriate functional capability of parameters and equipment within containment that are required for mitigation of design basis accidents during refueling operations (moving fuel to and from the reactor core). All TS in Section 3.9 are being proposed for deletion, as identified in the table above. TS 3.9.1, "Boron Concentration," places limits on the boron concentrations of the RCS and the refueling canal to ensure that the reactor remains subcritical during MODE 6. Refueling boron concentration is the soluble boron concentration in the filled portions of the RCS, the refueling canal, and the refueling cavity that are hydraulically coupled to the reactor core during refueling . The refueling boron concentration limit is specified in the COLR. The specified boron concentration is controlled by plant procedures to maintain an overall core reactivity of ke11::; 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) . TS 3.9.1 is applicable in MODE 6. TS 3.9.2, "Unborated Water Source Isolation Valves," requires flow paths to the RCS from unborated water sources be isolated to prevent unplanned boron dilution during MODE 6 and thus avoid a reduction in SOM. TS 3.9.2 contains requirements that during MODE 6 operations, all isolation valves for reactor makeup water sources containing unborated water that are connected to the RCS must be closed to prevent unplanned boron dilution of the reactor coolant. TS 3.9.2 is applicable in MODE 6. TS 3.9.3, "Nuclear Instrumentation," requires that two source range flux monitors be operable to ensure that redundant monitorinq capability is available to detect chanqes in core reactivity. The two source ranqe flux Page 62 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 3.9 - Refueling Operations monitors are required to provide a signal to alert the operator to unexpected changes in core reactivity such as a boron dilution accident or an improperly loaded fuel assembly . TS 3.9.3 is applicable in MODE 6. TS 3.9.4, "Containment Penetrations," specifies requirements for containment closure (i.e., all potential escape paths are filtered, closed, or capable of being closed) during movement of recently irradiated fuel assemblies within containment. This LCO limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed . TS 3.9.4 is applicable during movement of recently irradiated fuel assemblies within containment. TS 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," requires one RHR loop to be operable and in operation in MODE 6 with the water level ~ 23 ft above the top of reactor vessel flange. The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the RCS to provide mixing of borated coolant and to prevent boron stratification . Only one RHR loop is required for decay heat removal in MODE 6, with the water level ~ 23 ft above the top of the reactor vessel flange because the volume of water above the reactor vessel flange provides backup decay heat removal capability. One RHR loop is required to be in operation and operable to provide removal of decay heat; mixing of borated coolant to minimize the possibility of criticality; and indication of reactor temperature. TS 3.9.5 is applicable in MODE 6 with the water level~ 23 ft above the top of reactor vessel flange. TS 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level," requires two RHR loops to be operable and one RHR loop to be in operation in MODE 6 with the water level< 23 ft above the top of reactor vessel flange. The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the RCS to provide mixing of borated coolant and to prevent boron stratification. Both RHR loops are required to be operable in MODE 6, with the water level< 23 ft above the top of the reactor vessel flange . In addition , one RHR loop is required to be in operation in order to provide removal of decay heat; mixing of borated coolant to minimize the possibility of criticality; and indication of reactor temperature. TS 3.9.6 is applicable in MODE 6 with the water level < 23 ft above the top of reactor vessel flange. TS 3.9.7, "Refueling Cavity Water Level," specifies a minimum refueling cavity water level of 23 ft above the reactor vessel flange that is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits. The movement of irradiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange . This requirement ensures a sufficient level of water is maintained in the refueling cavity to retain iodine fission product activity resulting from a fuel handling accident in containment. Sufficient iodine activity would be retained to limit offsite doses from the accident to within 10 CFR 50.67 limits. TS 3.9.7 is applicable during movement of irradiated fuel assemblies within containment. Summary The content of TS Section 3.9 is being proposed for deletion in its entirety. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Byron, the 10 CFR 50 licenses will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactor vessels in accordance with 10 CFR 50 .82(a)(2) . As a result, TS Section 3.9 will not apply with the reactor defueled. With TS Section 3.9 deleted in its entirety, the applicable Bases and surveillance section will also be deleted to reflect this change. Proposed License Conditions 2.C.(25) and 2 .C.(14) for Byron, Unit 1 and Unit 2, respectively, will prohibit movement of irradiated fuel in the SFP after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown. Page 63 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 4.0 - Design Features TS Being Deleted TS Being Retained 4.1 - Site 4.2 - Reactor Core 4.3 - Fuel Storage Current TS Pro12osed TS 4.1 .2 Exclusion Area Boundary (EAB) 4.1.2 Exclusian Area Baundary (E/\B)De/eted The EAB shall not be less than 1460 ft (445 +Re E:AB st:iall net se less tt:ian ~ 4eQ ft (44a meters) from the outer containment wall. R'leteFS) fFaR'I the auter cantainR'lent wall. 4.1 .3 Low Population Zone (LPZ) 4.1.3 Law Papulatian Zane (LPZ)Oe/eted The LPZ shall be a 3.0 mi (4828 meter) radius +Re LP~ st:iall sea ;3.Q R'li (4828 R'leter) radius measured from the midpoint between the two R'leasured frnR'I tt:ie R'lidpaint setween tt:ie r.va reactors . reactarsDe/eted Basis TS Section 4.0, "Design Features," contains a brief description of the Byron location, description and requirements for the Byron reactor cores , and a description of and requirements for fuel storage at Byron. TS 4.1, "Site," contains the Site Location, the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ). In accordance to 10 CFR 50.82(a)(2), the facility licenses for Byron will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. With the reactors permanently shutdown and defueled , the locations of the reactors relative to the LPZ or the EAB to the outer containment wall are inconsequential. Therefore, TS 4.1 will be revised to remove the references to the EAB and LPZ. TS 4.2, "Reactor Core," contains descriptions of and requirements for fuel assemblies and control rod assemblies in the Byron reactor core during operation of the units. In accordance with 10 CFR 50 .82(a)(2), the facility licenses for Byron will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors . With the reactor permanently shutdown and defueled, the operational reactor cores, as described in TS 4.2, will no longer exist. Therefore, TS 4.2 will be deleted in its entirety. Because the TS will not be renumbered, a markup is provided in Attachment 2 to identify this section as deleted. TS 4.3 will remain applicable with the reactors permanently defueled. As such, this TS section is being retained to reflect a permanently defueled condition. Summarv TS 4.1 is being retained and revised. Retaining TS 4.1.1 ensures appropriate requirements for the associated design features . The information contained in TS 4.1.2 and TS 4.1.3 is inconsequential with the reactors permanently shutdown and defueled; therefore, these TS are proposed for deletion. TS 4.2 is proposed for deletion is in its entirety, as identified in the table above . TS 4.3 will remain applicable with the reactors permanently defueled . As such, this TS section is being retained to reflect a permanently defueled condition . Page 64 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS Section 5.0 - Administrative Controls TS Being Deleted TS Being Retained 5.1 - Responsibility 5.2 - Orqanization 5.3 - Facility Staff Qualifications 5.4 - Procedures 5.5 - Programs and Manuals 5.6 - Reportinq Requirements 5.7 - Hiqh Radiation Area Basis The existing TS Section 5.0, "Administrative Controls," contains provisions relating to organization and management, procedures, programs, reporting requirements, high radiation area necessary to ensure operation of the facility in a safe manner. There are no TS Bases associated with TS 5.0. TS 5.1 - Responsibility TS 5.1 is being retained into the POTS. Exelon proposed changes to the administrative controls in TS 5.1 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3). TS 5.2 - Organization TS 5.2 is being retained into the POTS. Exelon proposed changes to the administrative controls in TS 5.2 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3). TS 5.3 - Facility Staff Qualifications TS 5.3 is being retained into the POTS. Exelon proposed changes to the administrative controls in TS 5.3 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3). TS 5.4 - Procedures TS 5.4 is being retained into the POTS. Exelon proposed changes to the administrative controls in TS 5.4 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3). TS 5.5.2 - Prima[V Coolant Sources Outside This program was established to provide controls to Containment minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. It addresses the recirculation portions of the Containment Spray, Safety Injection , Chemical and Volume Control , and Residual Heat Removal. As previously discussed, the TS requirements for these systems were proposed for deletion. Once Byron is permanently shutdown and defueled , there will no longer be any transient or accident conditions associated with primary coolant Page 65 of 79

ATTACHMENT 1 Evaluation of Proposed Changes sources. Therefore, TS 5.5.2 does not apply in the permanently shutdown and defueled condition and will not be retained. TS 5.5.4 - Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program is contained in the ODCM. The term "each unit" is being changed to "the facility" to align with a permanently shutdown and defueled condition, which is consistent with other changes described above. TS 5.5.5 - Com1::1onent C:tclic or Transient Limit This program provides controls to track the cyclic and transient occurrences to ensure that components are maintained within the design limits. The program will not be retained in the POTS, because the Byron 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessels once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC. The reactor vessels will no longer be subjected to cycles or transients after permanent shutdown. TS 5.5.6 - Pre-Stressed Concrete Containment This program provides controls for monitoring any Tendon Surveillance Program tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. TS 5.5.6 is being deleted because the program pertains only to the containment structure that does not apply in a permanently defueled condition. TS 5.5.7 - Reactor Coolant Pumg Fl~heel lnsgection This program provides for the inspection of each Program reactor coolant pump flywheel in general conformance with the recommendations of Regulatory Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975. TS 5.5.7 is being deleted because the Reactor Coolant Pump Flywheel Inspection Program pertains only to reactor support systems that do not apply in a permanently defueled condition. TS 5.5.9 - Steam Generator (SG} Program This program is established and implemented to ensure that SG tube integrity is maintained. TS 5.5.9 is being deleted because the Steam Generator Program pertains only to reactor support systems that do not apply in a permanently defueled condition. TS 5.5.10 - Seconda[Y Water Chemistr:t Program This program is established and implemented to inhibit SG tube degradation . TS 5.5.10 is being deleted because the Secondary Water Chemistry Program pertains only to reactor support systems that do not apply in a permanently defueled condition . Page 66 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS 5.5.11 - Ventilation Filter Testing Program (VFTP) This program was established to implement the required testing of the Engineered Safety Feature (ESF) filter ventilation systems. This program will not be retained in the PDTS, because the VFTP is no longer required in a permanently shutdown and defueled condition . As previously discussed, TS 3.7 that provided the operability requirements associated with ESF filter ventilation systems are proposed to be deleted. TS 5.5.13 - Diesel Fuel Oil Testing Program The Diesel Fuel Oil Testing Program was established to implement the required testing of both new and stored fuel oil for the DGs. This program is proposed for elimination from the PDTS since the DGs will not perform any safety function in the permanently shutdown and defueled facility. The TS associated with the DGs and the diesel fuel oil subsystem (TS 3.8.1, TS 3.8.2, and TS 3.8.3) are proposed for removal from the PDTS as described above. TS 5.5.14 - Technical Specifications (TS) Bases The Technical Specifications Bases Control Program Control Program is being modified to reflect that once the facility is permanently defueled the title of the UFSAR will be revised to DSAR. TS 5.5.15 - Safety Function Determination Program This program was established to ensure loss of safety (SFDP) function is detected and appropriate actions taken . The SFDP is proposed for elimination since the LCOs remaining in the PDTS do not rely on the operability of any active equipment or systems to satisfy the LCO. Because 10 CFR 50 .82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, there is no longer a need for redundant systems. Therefore, the requirements of the SFDP, which directs cross train checks of multiple and redundant safety systems, no longer apply. Additionally, the SFDP is invoked by LCO 3.0.6, which is being deleted in its entirety as previously discussed. Therefore, this specification does not apply in the permanently shutdown and defueled condition . TS 5.5.16 - Containment Leakage Rate Testing This program was established to implement the Program leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50 Appendix J, Option B, as modified by exemptions. This program will not be retained in the PDTS, because the Containment Leakage Rate Testing Program pertains only to reactor support systems that are not needed in a permanently defueled condition . The requirements in TS 3.6.1 for containment systems are being deleted as described above. Therefore, this specification does not apply in the permanently shutdown and defueled condition. Page 67 of 79

ATTACHMENT 1 Evaluation of Proposed Changes TS 5.5.17 - Batte!}'. Monitoring and Maintenance This program was established to provide for battery Program restoration and maintenance. As discussed above, TS Section 3.8, including all requirements for DC sources and battery parameters are proposed for deletion in their entirety. Therefore, the requirements for the maintenance, testing, and replacement of batteries described in this specification are similarly unnecessary and are proposed for deletion following the establishment of a permanently shutdown and defueled condition for Byron. TS 5.5.18 - Control Room Envelope Habitability This program was established and implemented to Program ensure that the Control Room Envelope (CRE) habitability was maintained such that, with an operable VC Filtration System , the occupants of the CRE can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release or a smoke challenge. Following permanent shutdown and defueling, and through the imposition of the proposed License Conditions 2.C.(25) and 2.C.(14) for Byron, Units 1 and 2, respectively , the analysis of the FHA demonstrates that the CRE is not required for providing airborne radiological protection for the control room operator. Moreover, the majority of the controls associated with the handling and storage of irradiated fuel for Byron are located outside the control room. As such, the need for the operator to occupy the CRE is diminished following the establishment of the permanent shutdown and defueled condition. Additionally, as previously discussed, TS 3.3.7 is not proposed to be included in the POTS; thus, Technical Specification 5.5.18 is proposed for deletion. TS 5.5.20 - Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT). The Technical Specifications proposed for retention in the POTS do not contain any RICT allowances. Therefore, there is no need to maintain this program to RICTs and it can be eliminated. Thus, Technical Specification 5.5.20 is proposed for deletion. TS 5.6 .2 - Annual Radiological Environmental TS 5.6.2 is being retained in the POTS with minor Operating Report editorial changes. The NOTE regarding the ability to make a single submittal for a multiple unit station is proposed for deletion. Once Byron has ceased operations, Units 1 and 2 will be referred to as the facility. Additionally, the term "operation of the" is deleted to reflect the change in status regarding Byron. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Byron , the 10 CFR 50 licenses will no longer authorize operation of the Page 68 of 79

ATTACHMENT 1 Evaluation of Proposed Changes reactors or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). TS 5.6 .3 - Radioactive Effluent Release Report TS 5.6.3 is being retained in the POTS with minor editorial changes. The NOTE regarding the ability to make a single submittal for a multiple unit station is proposed for deletion . Once Byron has ceased operations, Units 1 and 2 will be referred to as the facility. Additionally, the term "operation of the" is deleted to reflect the change in status regarding Byron. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Byron, the 10 CFR 50 licenses will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). TS 5.6.5 - Core Operating Limits Report (COLR} According to TS Section 5.6 .5, the core operating limits shall be established prior to each reload cycle , or prior to any remaining portion of a reload cycle , and shall be documented in the COLR to ensure that all limits of the safety analysis are met. The specific limits are associated with operation the reactor core. This reporting requirement is not proposed for inclusion in the POTS , because the Byron 10 CFR 50 licenses will prohibit operation of the reactors or emplacement or retention of fuel in the reactor vessels once the certifications required by 10 CFR 50.82(a)(1) have been submitted . Thus, the COLR does not apply in the permanently shutdown and defueled condition . TS 5.6 .6 - Reactor Coolant System (RCS} This specification being deleted because the PTLR PRESSURE AND TEMPERATURE LIMITS REPORT pertains only to an activity that does not apply in a (PTLR} permanently defueled condition . The two TS that reference the PTRL (listed in TS 5.6.6.a) are being deleted as described in preceding sections. Therefore, the need for this report no longer exists. TS 5.6 .7 - Post Accident Monitoring Report This report is required by Condition C or G of TS 3.3.3 . The report outlines the preplanned alternate method of monitoring , the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to an operable status. This reporting requi rement will not be retained in the POTS because the Byron 10 CFR 50 licenses will prohibit operation of the reactors or emplacement or retention of fuel in the reactor vessels once the certifications required by 10 CFR 50 .82(a)(1) have been submitted to the NRC. As discussed above, the requirements in TS 3.3.3 regarding PAM instrumentation are proposed to be deleted : therefore , the PAM instrument report will not be required in the Page 69 of 79

ATTACHMENT 1 Evaluation of Proposed Changes permanently shutdown and defueled condition and is proposed for deletion. TS 5.6.8 - Tendon Surveillance Regort This specification establishes a requirement to submit a report in the lnservice Inspection Summary report for any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program. This report is being deleted because it pertains to activities that do not apply in a permanently defueled condition. TS 5.6.9 - Steam Generator (SG} Tube lnsgection This specification establishes a requirement to submit Rego rt a report within 180 days of the initial entry into Mode 4 following completion of an inspection performed in accordance with TS 5.5.9, "Steam Generator (SG) Program ." As described above, TS 5.5.9 is proposed for deletion. Therefore, TS 5.6.9 will no longer be required and is also proposed for deletion. provides the existing TS pages for Byron, marked up to show the proposed changes. Proposed changes to the TS Bases addressing the proposed changes to the relevant TS are provided for information in Attachment 4.

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. Exelon has determined that the proposed changes do not require any exemptions or relief from regulatory requirements. 10 CFR 50.82. Termination of License The 10 CFR 50.82(a)(1) paragraph requires that when a licensee has determined to permanently cease operations, the licensee shall , within 30 days, submit a written certification to the NRC, consistent with the requirements of 10 CFR 50.4(b)(8), and once fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC that meets the requirements of 10 CFR 50.4(b)(9). The 10 CFR 50.82(a)(2) paragraph states, "Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel." By letter dated September 2, 2020 (Reference 1), Exelon provided formal notification to the NRC pursuant to 10 CFR 50.4(b)(8) and 10 CFR 50.82(a)(1)(i) that it would permanently cease operations at Byron on or before September 30, 2021. Page 70 of 79

ATTACHMENT 1 Evaluation of Proposed Changes Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessels are submitted for Byron to the NRG pursuant to 10 CFR 50.82(a)(1 ), NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the reactors or emplacement of fuel into the reactor vessels under the 10 CFR 50 licenses. As a result, Byron will be authorized only to possess special nuclear material. 10 CFR 50.36. Technical Specifications In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity." (Statement of Consideration, Technical Specification for Facility Licenses; Safety Analysis Reports," 33 FR 18610 (December 17, 1968)) Pursuant to 10 CFR 50.36, TS are required to include items in the following five categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls . However, the rule does not specify the particular requirements to be included in a facilities' TS. These criteria, which were subsequently codified in changes to 10 CFR 50.36 (60 FR 36953), also pertain to the TS requirements for safe storage of irradiated fuel. A general discussion of these considerations is provided below to address the existing LCOs. Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for "installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the reactor coolant pressure boundary." Since no fuel will be present in the reactor or RCS at Byron in the permanently shutdown and defueled condition, this criterion is not applicable. Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a "process variable , design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the OBA and transient analyses, and which are monitored and controlled during power operation. While this criterion was developed for operating reactors, there are some DBAs which continue to apply to a facility authorized only to handle, store, and possess irradiated fuel. The scope of DBAs applicable to a facility with a reactor that is permanently shutdown and defueled is markedly reduced from those postulated for an operating plant. The applicable DBAs for Byron in the permanently defueled condition are discussed in more detail within this license amendment request. Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS LCOs must be established for SSCs that are part of the primary success path and which function or actuate to mitigate a OBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into the TS only those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success Page 71 of 79

ATTACHMENT 1 Evaluation of Proposed Changes path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria. While there are no transients that will continue to apply to Byron , there are still DBAs that will continue to apply to a facility authorized only to handle, store, and possess irradiated fuel. The scope of DBAs applicable to a facility with a reactor that is permanently shutdown and defueled is markedly reduced from those postulated for an operating plant. The scope of DBAs that will be applicable to Byron is discussed in more detail within this license amendment request. Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS LCOs must be established for SSCs that operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. All of the accident sequences that previously dominated risk at Byron will no longer be applicable after the reactor is in the permanently shutdown and defueled condition. Addressing administrative controls, 10 CFR 50.36(c)(5) states that they" ...are the provisions relating to organ ization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." This license amendment request is proposing changes to the Administrative Controls section, with conforming changes proposed to additional sections, consistent with the pending decommissioning status of the plant. This request applies the principles identified in 10 CFR 50.36(c)(6), "Decommissioning," for a facility which has submitted certifications required by 50.82(a)(1) and proposes changes to the Administrative Controls appropriate for the Byron permanently defueled condition. As 10 CFR 50.36(c)(6) states, this type of change should be considered on a case-by-case basis. The 10 CFR 50 .36(c)(6), "Decommissioning," provisions apply only to nuclear power reactor facilities that have submitted the certifications required by 10 CFR 50.82(a)(1 ). For such facilities, TS involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation ; surveillance requirements ; design features; and administrative controls will be developed on a case-by-case basis. This proposed amendment deletes the portions of the Byron TS that are no longer applicable to a permanently defueled facility while modifying the remaining portions to correspond to the permanently shutdown and defueled condition. 10 CFR 50.46. Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors The 10 CFR 50.46(a)(1 )(i) paragraph states, "This section does not apply to a nuclear power reactor facility for which the certifications required under 10 CFR 50 .82(a)(1) have been submitted." Page 72 of 79

ATTACHMENT 1 Evaluation of Proposed Changes 10 CFR 50.48(f), Fire Protection During Decommissioning The 10 CFR 50.48(f) paragraph states, in part, that: "Licensees that have submitted the certifications required under 10 CFR 50.82(a)(1) shall maintain a fire protection program to address the potential for fires that could cause the release or spread of radioactive materials (i.e., that could result in a radiological hazard) .. . (1) The objectives of the fire protection program are to - (i) Reasonably prevent these fires from occurring; (ii) Rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and (iii) Ensure that the risk of fire-induced radiological hazards to the public environment and plant personnel is minimized. (2) The licensee shall assess the fire protection program on a regular basis. The licensee shall revise the plan as appropriate throughout the various stages of facility decommissioning. (3) The licensee may make changes to the fire protection program without NRC approval if these changes do not reduce the effectiveness of fire protection for facilities, systems, and equipment that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities." 10 CFR 50.51. Continuation of License The 10 CFR 50.51 (b) paragraph states, "Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated . During such period of continued effectiveness , the licensee shall - (1) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including , where applicable, the storage, control and maintenance of the spent fuel , in a safe condition, and (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR part 50 license for the facility." 10 CFR 50.62. Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants The 10 CFR 50.62(a) paragraph states, "The requirements of this section apply to all commercial lightwater-cooled nuclear power plants, other than nuclear power reactor facilities for which the certifications required under§ 50.82(a)(1) have been submitted." Page 73 of 79

ATTACHMENT 1 Evaluation of Proposed Changes 10 CFR 50.67. Accident source term. (a) Applicability. The requirements of this section apply to all holders of operating licenses issued prior to January 10, 1997, and holders of renewed licenses under part 54 of this chapter whose initial operating license was issued prior to January 10, 1997, who seek to revise the current accident source term used in their design basis radiological analyses. (b) Requirements. ( 1) A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under

     § 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

(2) The NRG may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that: (i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE). (ii) An individual located at any point on the outer boundary of the low population zone , who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE). (iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident." 10 CFR 50. Appendix A. General Design Criteria (GDC) for Nuclear Power Plants Section 3.1 of the Byron UFSAR states, "The intent of the design of the Byron/Braidwood Stations is to conform to the Licensee's interpretation of the intent of Appendix A to 10 CFR 50." The Byron UFSAR details the interpretation of conformance , concluding that Byron fully satisfies and is in compliance with the NRG General Design Criteria. Design Basis Accidents (OBAs) Chapter 15 of the Byron UFSAR describes the OBA scenarios that are applicable during plant operations. After certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessels are submitted to the NRG in accordance with 10 CFR 50.82(a)(1) for Byron, the 10 CFR 50 licenses will no longer permit operation of the reactors or emplacement of fuel in the reactor vessels in accordance with 10 CFR 50 .82(a)(2). With the reactors in a permanently shutdown and defueled condition, the facility mission changes . The primary mission is now the safe storage and handling of irradiated fuel. In this condition , the spectrum of credible accidents is much smaller than for an operational plant. Therefore, most of the accident scenarios postulated in U FSAR Chapter 15 will no longer be applicable after Byron is in the permanently defueled condition. The events that remain applicable to Byron , Units 1 and 2, in the permanently shutdown and defueled condition are the Fuel Handling Accident Page 74 of 79

ATTACHMENT 1 Evaluation of Proposed Changes (FHA) within the SFP, Gas Waste System Leak or Failure, Radioactive Liquid Waste System Leak or Failure, Postulated Radioactive Release Due to Liquid Tank Failure, and Spent Fuel Cask Drop Accident. The FHA within containment is no longer applicable with the reactor permanently defueled. 3.2 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) requests amendments to the Renewed Facility Operating Licenses (RFOLs) and Appendix A, Technical Specifications (TS), of RFOL Nos. NPF-37 and NPF-66 for Byron Station , Units 1 and 2 (Byron). The proposed amendments revise the RFOLs and TS consistent with the permanent cessation of operation and defueling of the reactors. The revised RFOLs and TS will be identified as the Byron Permanently Defueled Technical Specifications (POTS). Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessels are submitted to the NRC pursuant to 10 CFR 50.82(a)(1 ), NRC regulations stipulated in 10 CFR 50 .82(a)(2) will no longer authorize operation of the reactors or emplacement of fuel into the reactor vessels under the 10 CFR 50 licenses. Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No. The proposed changes would not take effect until Exelon has submitted the certifications required by 10 CFR 50.82(a)(1) that Byron, Units 1 and 2, have permanently ceased operation and have entered a permanently defueled condition . Because the 10 CFR 50 licenses for Byron will no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible . Byron's accident analyses are contained in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). The remaining postulated design basis accident (OBA) events that could potentially occur at Byron in a permanently defueled condition would be a Fuel Handling Accident (FHA) in the Fuel Handling Building , gas waste system leak or failure , radioactive liquid waste system leak or failure , postulated radioactive release due to liquid tank failure, and spent fuel cask drop accident. The FHA analysis in the Fuel Handling Building for Byron, Units 1 and 2, shows that after 70 days of decay time after the reactors have shutdown and provided the SFP water level requirement of LCO 3.7.14 is met, the dose consequences are acceptable without relying on active SSCs to remain functional for accident mitigation during and following the event. To preclude an FHA with unacceptable radiological consequences, Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessels until 70 days after permanent shutdown. The remaining DBAs that support the permanently shutdown and defueled condition do not rely on any active safety systems for mitigation. Page 75 of 79

ATTACHMENT 1 Evaluation of Proposed Changes The probability of occurrence of previously evaluated accidents is not increased because safe storage and handling of irradiated fuel will be the only operations performed, and these activities are bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible with permanently defueled reactors. This significantly reduces the scope of applicable accidents. The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shutdown and defueled Byron facility has no impact on the remaining applicable OBAs. The removal of LCOs and associated SRs that relate only to the operation of the nuclear reactors or the prevention , diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable OBAs previously evaluated because these OBAs will no longer be applicable in the permanently defueled condition. Therefore, the proposed changes do not involve a significant increase in the probability or consequence of an accident previously evaluated .

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

Response: No. The proposed changes to delete or modify the RFOLs, additional conditions, and TS have no impact on facility SSCs affecting the safe storage and handling of irradiated fuel. The removal of TS that are related only to the operation of the nuclear reactors, or only to the prevention, diagnosis, or mitigation of reactor-related transients and accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactors will be permanently shutdown and defueled. The proposed modifications and deletion of requirements in the ByronRFOLs, additional conditions, and TS do not affect systems credited in the accident analysis for the remaining OBAs. The proposed RFOLs and POTS will continue to require proper control and monitoring of safety significant parameters and activities. The TS regarding SFP water level , boron concentration, and spent fuel assembly storage are retained to preserve the current requirements for safe storage of irradiated fuel. The proposed changes do not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers in support of maintaining the facility in a permanently shutdown and defueled condition. Since safe storage and handling of irradiated fuel will be the only operations allowed , and these activities are bounded by the existing analyses, the proposed changes do not create the possibility of a new or different kind of accident. The proposed changes do not create new failure modes. The proposed changes do not involve a physical alteration of the plant, and no new or different kind of equipment will be installed . Consequently, there are no new initiators that could result in a new or different kind of accident. Page 76 of 79

ATTACHMENT 1 Evaluation of Proposed Changes Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated .

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

Response: No. The proposed changes include deleting or modifying RFOLs, additional conditions, and TS once Byron , Units 1 and 2, have permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 licenses for Byron will no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels following submittal of the certifications required by 10 CFR 50 .82(a)(1 ). Because the 10 CFR 50 licenses for Byron will no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels , the occurrence of postulated accidents associated with reactor operation will no longer be credible. The only remaining postulated OBA events that could potentially occur at Byron in a permanently defueled condition are an FHA in the FHB, gas waste system leak or failure, radioactive liquid waste system leak or failure, postulated radioactive release due to liquid tank failure, and spent fuel cask drop accident. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses as discussed in this license amendment request. The proposed changes are limited to those portions of the RFOLs, additional conditions, technical specifications, and licensing basis that are not related to the safe storage and handling of irradiated fuel. The requirements proposed to be revised or deleted from the RFOLs, additional conditions, and TS are not credited in the updated applicable accident analyses for the remaining applicable postulated accidents, and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs involving the reactors will no longer be possible because the reactors will be permanently shutdown and defueled, and operation of Byron reactors will no longer be authorized . To preclude an FHA with unacceptable radiological consequences, Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 70 days after permanent shutdown. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. Based on the above, Exelon concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified . 3.3 Conclusion In conclusion, based on the considerations discussed above, 1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 2) such activities will be conducted in compliance with the Commission's regulations, Page 77 of 79

ATTACHMENT 1 Evaluation of Proposed Changes 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.

4.0 ENVIRONMENTAL CONSIDERATION

Exelon has evaluated the proposed amendments for environmental considerations. The review has determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). In addition, the proposed changes involve changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

5.0 REFERENCES

1. Letter from J. Bradley Fewell (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Certification of Permanent Cessation of Power Operations for Byron Station, Units 1 and 2," dated September 2, 2020 (NRC Accession No. ML20246G613)

2. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Request for Approval of Certified Fuel Handler Training and Retraining Program," for Byron Station, Units 1 and 2, and Dresden Nuclear Power Station, Units 1, 2 and 3, dated September 24, 2020 (NRC Accession No. ML20269A233)

3. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "License Amendment Request - Proposed Changed to Technical Specifications Section 1.1, 'Definitions,' and 5.0, 'Administrative Controls,' for Permanently Defueled Condition," dated September 24, 2020 (NRC Accession No. ML20269A401)

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

5. Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," dated March 7, 2006 Page 78 of 79

ATTACHMENT 1 Evaluation of Proposed Changes

6. Letter from John G. Lamb (U .S. Nuclear Regulatory Commission) to Bryan C. Hanson (Exelon Generation Company, LLC}, "Oyster Creek Nuclear Generating Station and Independent Fuel Storage Installation - Review and Acceptance of Changes RE:

Decommissioning Quality Assurance Program (EPID L-2017-LLQ-0003)," dated June 27, 2018 (NRG Accession No. ML18165A136) Page 79 of 79

ATTACHMENT 2 Markup of Renewed Facility Operating License Nos. NPF-37 and NPF-66 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 MARKED UP UNIT 1 RENEWED FACILITY OPERATING LICENSE PAGES Pages 1-8, INSERT MARKED UP UNIT 2 RENEWED FACILITY OPERATING LICENSE PAGES Pages 1-7, INSERT

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 RENEWED FACILITY OPERATIP~G LICENSE Renewed License No. NPF-37

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

A. The application for a renewed license filed by the applicant* complies with the standards and requirements of the Atomic Energy Act of 1954, as amend (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. CeRstH:1stieR ef tl:ie BymR ~tatieR , URit ~Je . 1 (tl:ie fasility) l:ias eeeR s1:18staRtially seFR13leteEI iR seRferFRity witl:i CeRstr1:1stieR PerFRit Me . CPPR 1JG a REI tl:ie a1313lisatieR , as aFReREleEI , tl:ie 13revisieRs ef tl:ie Ast, aREI tl:ie re§l1:1latieRs ef tl:ie CeFRFRissieR ; pbe maintained I C. The facility will e13erate in conformity with the application , as amended , the provisions of the Act, and the regulations of the Commission (except as exempted from compliance in Section 2.D below); D. There is reasonable assurance: (i) that the activities authorized by this renewed e13eratiR§l license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission 's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D below) ; E. Exelon Generation Company, LLC is technically qualified to engage in the activities authorized by this renewed license in accordance with the Commission's regulations set forth in 10 CFR Chapter I; F. Exelon Generation Company, LLC has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements, " of the Commission 's regulations;

The Nuclear Regulatory Commission approved the transfer of the license from Commonwealth Edison Company to Exelon Generation Company, LLC on August 3, 2000 .

Renewed License No. NPF-37

G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; H. After weighing the environmental , economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Renewed Facility 013eratin§ License No. NPF-37, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. Tl'le resei13t, 13ossession, anEI 1:1se of so1:1rse , l3y13roEl1:1st anEI s13esial n1:1slear

        ....--------. 1f n::iaterial as a1:1tl:ierizeel lay tl:iis reneweel Ii sense will lae in asserelanse witl:i tl:ie
        ~ Gen::in::iissien 's F8§1:llatiens in 19 Gl=R Parts aG, 4g anel 79.

J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging E11:1rin§ tl:ie 13erieel ef 8}<teneleel e13eratien on the for those Unit 1 and functionality of structures and components that have been identified to require Common Plant review under 10 CFR 54.21 (a)(1 ), and (2) time-limited aging analyses that have been identified to require review1 ~nder 10 CFR 54.21 (c), such that there is systems , structures, reasonable assurance that the a(,tivities authorized by this renewed license will and components continue to be conducted in accordance with the current licensing basis, as that remain in defined in 10 CFR 54 .3, for the facility, and that any changes made to the service beyond facility's current licensing basis in order to comply with 10 CFR 54 .29(a) are in October 30, 2024, accordance with the Act and the Commission 's regulations.

2. Facility 013eratin§ License No. NPF-37, dated February 14, 1985, as amended, is hereby superseded by Renewed Facility 013eratin§ License No. NPF-37, issued to Exelon Generation Company, LLC (the licensee) to read as follows: £ permanently defueled I A. This renewed license applies to the Byron Station, Unit No . 1, a pressurized water nuclear reactor, and associated equipment (the facility), owned by Exelon Generation Company, LLC . The facility is located in north central Illinois within Rockvale Township, Ogle County, Illinois and is described in the licensee's "Updated Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report, as supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Exelon Generation Company, LLC : ( 1) Pursuant to Section 103 of the Act and 10 CFR Part 50 to possess;-l::!Se

                  ~                ~anel e13erate the facility at the designated location in accordance with the
                  ~                  *procedures and limita        1 ns set forth in this renewed license; as required for irradiated fuel storage                                       Renewed License No. NPF-37

that was used (2) Pursuant to the Act and 10 CFR art 70, to Feeei'v'e, possess aAel 1:1se at any time special nuclear material as reactor fuel, in accordance with the limitations for storage aRel aFRe1:1Rts Feei1:1ireel fer reaster e13eratieR , as described in the Updated Final Safety Analysis Report, as supplemented and amended; and to possess any (3) Pursuant to the Act and 10 CFR Parts 30, 40 byproduct, source and use at any time any byproduct, source aftfl;-S~618rf.-A.~eaF-ff.OO:H=h:* as sealed Re1:1treR se1:1rses fer reaster start1:113 , sealed sources for reaster and special nuclear iAstr1:1FReRtatieA aRel radiation monitoring equipment calibration, aRel as material as sealed fissieR detesters iR aFR01:1Ats as reei1:1ired ; neutron sources previously used for (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive , possess, reactor startup or and use in amounts as required any byproduct, source or special nuclear reactor material without restriction to chemical or physical form , for sample instrumentation; and analysis or instrument calibration or associated with radioactive apparatus or components; and fission detectors (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate , such byproduct and special nuclear materials as FRay be produced by the operation of the facility. ~ C. The 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) MmciFRl:lFR Pewer Level ~ Tl::ie liseRsee is a1:1tl::ierizeet te e13erate tl::ie fasility at reaster sere 13ewer levels Aet iA mwess ef 364 § ffle§awatts tl=lerfflal (100 19ereeAt rateel 13ewer) iR asserelaRse witl=i tl=ie seRelitieRs s13esifieel l::iereiR . Permanently (2) Technical Specifications

                    ----~

Defueled The echnical Speciti* tions contained in Appendix A as revised through Amendment No. ~ and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall e13erate the facility in accordance with the Technical Specifications and th nvironmental Protection Plan . maintain Permanently Defueled (3) Deleted. (4) Deleted. Renewed License No. NPF-37 Amendment No. ~

(5) Deleted. (6) The licensee shall irnplernent and rnaintain in effect all provisioAs of the ....------..:.. * :1fapproved fire protection progra1¥1 as described in the licensee's Fire ~ Protection Report, and as ap13roved in the SER dated February 1987 through Supplernent j')Jo. 8, subject to the following provision : The licensee rnay 1¥1ake cl:langes to tl:le a13wovea fiFe proteetion wograrn without prior approval of tl:le Cornrnission only if those ohanges would net adversely affeot tl:le ability to aol:lieve and rnaintain safe sl:lutdown in tl:le event of a fire. (7) Deleted (8) Deleted. (9) Deleted. (10) Deleted. ( 11) Deleted. (12) Deleted. (13) Deleted. (14) Deleted . (15) Deleted. (16) Deleted. (17) Additional Cenaitiens ~ Tl:le Additional Conditions oentained in Appendi>c C, as revised through ArnenelR9ent j')Ja. 198, are heresy ineer13erateel inte this reneweel license. Tl:le lioensee shall operate the faoility in aooorelanee with the Additional Conditions. Renewed License No. NPF-37 Amendment No. igg

(18) Exelon Generation Company, LLC shall provide the Director of the Office of Nuclear Reactor Regulation a copy of any application , at the time it is filed, to transfer (excluding grants of security interests or liens) from Exelon Generation Company, LLC to its direct or indirect parent, or to any other affiliated company, facil ities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company, LLC's consolidated net utility plant, as recorded on Exelon Generation Company, LLC's books of account. (19) Deleted. (20) Deleted. (21) Deleted. (22) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas: (a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (23) LiseRse ReRewal LiseRse CoRel itioRs

~ faj n1e iRferFRatioR iR tl::ie UP::£AR s1:11313leFReRt, Sl:IBFRitteel 131:1rs1:1aRt to 1G CFR §~ .21 (el) , as reYiseel el1:1riR§ tl::ie liseRse reRewa l a1313lisatioR review 13rosess, aRel as s1:11313leFReRteel 13y tl::ie CoFRFRitFReRts a1313lisal3le to ByroR UR it 1 iR A1313eReli)< A of tl"le Renewed License No. NPF-37 Amendment No. 2-14

                                                  "Safety e'v'alblatieR Re13eFI: Re lateel ts tl:le LiseRse ReRewal sf ByFeR StatieR , URits 1 aRel 2, aRel Bmielweeel StatieR , URits 1 aRel 2" (S E:R) elateel db!ly 201 e, is sell eetivel)' tl:le "bieeRse ReRewal UFeAR ebl1313 leffieRt. " Tl:l is ebl1313 leffieRt is ReRsefeFl:R 13ar=t sf tl:le UFeAR wl:l isl:l will 13e b113elateel iR asseFelaRse witl:l 10 GFR e0.71 (e). As Sb!SR , tl:le lieeRsee ffiay ffiake sl:laRges ts tl:le 13FO§Faffis aRel aetivities a1313lisal3 le to ByFoR URit 1 elessFil3eel iR tl:lis ebl1313leffieRt 13FO'v'ieleel tl:le liseRsee evalblates SblSR SRaRges 13blFSblaRt to tl::ie sriteria set fer=tl::i iR 1g GFR 90.99 a Rel etl::ierwise soffi13lies witl::i tl::ie FeEJbl irnffieRts iR tl::iat sestioR .

f9j Tl:l is LiseRse ReRewal UFe/\R ebl1313 leffieRt, as Fe'u'iseel 13eF LieeRse GeRelitieR 23(a) al3eve, eleseril3es eeFl:a iR 13regraffis te 13e iffi13leffieRteel aRel aetivities ts 13e seffi13 leteel 13Fier ts tl:le 13erieel sf e3cteReleel e13eratieR .

            .,.      Tl::ie liseRsee sl::ia ll iffi13leffieRt tl::iose Rew 13FegFaffis aRel eRRaReeffieRtS ta e>cisti Rg j3FO§FaffiS RS lateF tRaR A13Fi l 30 ,
                        ~
            ~           Tl:le li seRsee sl:la ll seffi13lete tl:lese aeti,*ities as Reteel iR tl:le GeffiffiitffieRts a1313lisa l3l e ts BymR URit 1 iR tl::i is Sb11313 leffieRt RO lateF tRaR A13Fil ao, 2024 SF tl::ie eRel of tl::ie last FefbleliRg ebltage 13FieF ts tl:le 13eFieel sf e)deReleel 013eFatieR , WRisl:leveF OSSblFS lateF.

a..,. Tl:le liseRsee SRall Ratify tl:le ~JRG iR WFitiRg witRiR ao ela;*s afteF Ra'u'iRg asseffi13lisl:leel iteffi (13)1 al3eve aRel iRelblele tl:le statbls sf tl:lese astivities tl:lat Rave 13eeR SF FeffiaiR ts 13e eeffi13 leteel iR iteffi (13)2 al3eve. (24) Adoption of 10 CFR 50.69. "Risk-informed categorization and treatment of structures. systems. and components for nuclear power plants" Exelon is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding , and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 ; as specified in the license amendment No. 204, dated October 22, 2018 . Renewed License No. NPF-37 Amendment No . ~

Exelon will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009 , as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process. Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a INSERT: seismic margins approach to a seismic probabilistic risk LICENSE assessment approach). CONDITION 2.C.(25) D. The facility requires no exemptions from the requirements of 10 CFR Part 50. E. Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualifications, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

                       "Byron Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 3, " submitted by letter dated May 17, 2006.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50 .90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 175 and modified by License Amendment No. 191 . F. Deleted G. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954 , as amended, to cover public liability claims . 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Renewed License No. NPF-37 Amendment No . ~

H. This renewed license is effective as of the date of issuance and sl:iall e~lJ3iFe at R.:iielRi§Rt OR Ootol3eF a1 ' 204 4. FOR THE NUCLEAR REGULATORY COMMISSION is effective until the Commission notifies the IRA/ licensee in writing that the license is terminated. William M. Dean, Director Office of Nuclear Reactor Regulation Appendices:

1. Appendix A- Technical Specifications (NUREG-1113)

2. Appendix B - Environmental Protection Plan

3. Appendix C - AelelitioRal GoRelitioRs ~

Date of Issuance: November 19, 2015 Renewed License No. NPF-37 Amendment No. 244

INSERT UNIT 1 RFOL LICENSE CONDITION 2.C.(25) (25) Handling of irradiated fuel in the spent fuel pool will not be permitted following implementation of the POTS until a minimum of 70 days following permanent shutdown.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 RENEWED FACILITY OPERATIPd G LICENSE Renewed License No. NPF-66

1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A. The application for a renewed license filed by the applicant* complies with the standards and requirements of the Atomic Energy Act of 1954, as amend (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. GeRstF1:1sti eR ef ByFeR ~ta ti eR , UA it 2 (tt::ie fasility) t::ias 13eeR s1:1l3staRtially =--------i 1fseFR13leteel iR seRfeFFRity witl::i GeRstF1:1stieR PeFFRit ~Je . GPPR 1d1 a Rel tl::ie ~ a1313 li satieR , as aFReReleel , tl::ie 13FevisieRs ef tl::ie Ast aRel tl::ie Fe§1:1 latieRs ef tl::ie GeFRFRissieR ; I//rlb

                                             . e main

d I
t a1ne C. The facility will e13ernte in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission (except as exempted from compliance in Section 2.D. below);

D. There is reasonable assurance: (i) that the activities authorized by this renewed e13e rntiR§ license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D . below); E. Exelon Generation Company, LLC is technically qualified to engage in the activities authorized by this renewed license in accordance with the Commission's regulations set forth in 10 CFR Chapter I; F. Exelon Generation Company, LLC has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations;

*The Nuclear Regulatory Commission approved the transfer of the license from Commonwealth Edison Company to Exelon Generation Company, LLC on August 3, 2000.

Renewed License No. NPF-66

G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; H. After weighing the environmental , economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Renewed Facility 013eratin§ License No. NPF-66, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B to Renewed License No. NPF-37, issued November 19, 2015, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. n1e Fesei13t, 13essessien, ans 1:1se ef se1:1Fse , l3y13Fe81:1st anel s13esial n1:1sleaF

         ......-------. 1' FAaterial as a1:1tRerizeel ey tRis renew eel lisense w ill ee in asserelanse w itR tRe
         ~ CeFAFAissien 's Fe§l:llatiens in 1Q ci;:R Parts JQ, 4Q anel 7G.

J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging el1:1Fin§ tRe 13eFieel ef exteneleel e13eratien on the functionality of structures and components that have been identified to require for those Unit 2 review under 10 CFR 54.21 (a)(1 ), and (2) time-limited aging analyses that have systems, structures, been identified to require review der 10 CFR 54 .21 (c), such that there is and components reasonable assurance that the a tivities authorized by this renewed license will that remain in continue to be conducted in accordance with the current licensing basis, as service beyond defined in 10 CFR 54 .3, for the facility, and that any changes made to the November 5, 2026, facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.

2. Facility 013eratin§ License No. NPF-66, dated January 30, 1987, as amended, is hereby superseded by Renewed Facility 013eratin§ License No. NPF-66, issued to Exelon Generation Company, LLC (the licensee) to read as follows: £_p_e_r_m_a_n_e_n_tl_y_d_e_f_u_e-le-d~I A. The renewed license applies to the Byron Station, Unit No. 2, a pressurized water reactor, and associated equipment (the facility), owned by Exelon Generation Company, LLC. The facility is located in north central Illinois within Rockvale Township, Ogle County, Illinois and is described in the licensee's Updated Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report, as supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Exelon Generation Company, LLC : ( 1) Pursuant to Section 103 of the Act and 10 CFR Part 50 to possess.,..t!Se

                   ~                 "/! anel e13erate the facility, at the designated location in accordance with the
                   ~                     procedures and limita i ns set forth in this renewed license; as required for irradiated                               Renewed License No. NPF-66 fuel storage

that was used (2) Pursuant to the Act and 10 CFR rt 70, to Feeei'v'e, possess aAel 1:1se at any time special nuclear material as reactor fuel, in accordance with the limitations for storage aRel aFRe1:1Rts Feei1:1ireel fer reaster e13eratieR , as described in the Updated Final Safety Analysis Report, as supplemented and amended; and to possess any (3) Pursuant to the Act and 10 CFR Parts 30, 40 byproduct, source and use at any time any byproduct, source 8ftfl:-S~618rf.-A.~eaF-ff.OO:H=h:* and special nuclear as sealed Re1:1treR se1:1rses fer reaster start1:113 , sealed sources for reaster iAstr1:1FReRtatieA aRel radiation monitoring equipment calibration, aRel as material as sealed fissieR detesters iA aFR01:1Ats as reei1:1ired ; neutron sources previously used for (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive , possess, reactor startup or and use in amounts as required any byproduct, source or special nuclear reactor material without restriction to chemical or physical form , for sample instrumentation; and analysis or instrument calibration or associated with radioactive apparatus fission detectors or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate , such byproduct and special nuclear materials as FRay be produced by the operation of the facility. ~ C. The 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) MmciFRl:lFR Power Level ~ Tl::ie liseRsee is a1:1tl::ierizeet te e13erate tl::ie fasility at reaster sere 13ewer levels Aet iA mwess ef 364 § ffle§awatts tl=lerfflal (100 19ereeAt rateel 13ewer) iR asserelaRse witl=i tl=ie seRelitieRs s13esifieel l::iereiR . Permanently (2 ) Technical Specifications D e f u e Ie d 1--------- Th echnical Specifications containe n Appendix A (NUREG-1113), as revised through Amendment No. ~. and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall e13erate the facility in accord~nce with the~ Jechnical Specifications and th nvironmental Protection Plan . p ti f d I maintain ermanen y 0 e ue 1e . Renewed License No. NPF-66 Amendment No. ~

(3) Deleted . (4) Deleted. (5) Deleted. (6) /\elelitioRal CoRelitioRs ~ Tl::ie AelelitioRal CoR elitioRs eoAtaiAeel iA A1313eAel i)( C, as Feviseel tl=iFobl §R

     /\FRORelFRORt ~Jo . 198, are l::ierel3y iR60FJ30Fateel iRtO tl::iis FOROWOel lieeRSO .

Tl::ie lieeRsee sl::iall 0130Fate tl::ie faeility iR aeeoFelaRee witl::i tl::ie /\elel itioRal CoRelitioRs. (7) Exelon Generation Company, LLC, shall provide to the Director of the Office of Nuclear Reactor Regulation a copy of any application , at the time it is filed , to transfer (excluding grants of security interests or liens) from Exelon Generation Company, LLC, to its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated net utility plant, as recorded on Exelon Generation Company, LLC's books of account. (8) Deleted . (9) Deleted. (10) Deleted. (11) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas: (a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures Renewed License No. NPF-66 Amendment No . ~

(c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders

( 12) LieeRse ReRewal LieeRse GoRelitioRs ~ fa1 nie iRformatioR iR tl:ie UFS/\R s1:11313lemeRt, Sl:IBmitteel J3l:IFSl:laRt to 10 GFR 94 .21 (el), as reviseel el1:1riR§ tl:ie lieeRse reRewal a1313li eatieR review 13reeess , aRel as s1:11313lemeRtea 13y tl:ie GommitmeRts a1313lieal3le to ByroR UR it 2 iR A1313eRElb< A of tl:ie "Safety Eval1:1atioR Re13ort Re lates to tl:ie LieeRse ReRewal of ByroR StatioR , URits 1 aRel 2, aRel BraiElvu'ooel StatioR , URits 1 aREl 2" (SER) Elates d1:1ly 201 § , is eell eetively tl:ie "LieeRse ReRewal UFS/\R 81:11313lemeRt. " Tl:iis S1:11313lemeRt is l:ieReefortl:i 13art of tl:i e UFSAR wl:iiel:i will so 1:113Elatea iR aeeerElaRee witl:i 10 GFR 90.71 (e) . As s1:1el:i , tl:ie li eeRsee may mal\e el:iaR§es to tl:ie 13ro§rams aREl aetivities a1313lieal3 le to ByroR URit 2 Eleseril3ea iR tl:iis S1:11313lemeRt 13rovieleel tl:ie lieeRsee eval1:Jates s1:1el:i el:iaR §OS 131,:1rs1:1aRt to tl:ie eriteria set fortl:i iR 1Q GFR 90.99 aREl etl:ierwise eem13lies witl:i tl:ie reeil:l iremeRts iR tl:iat seetioR .

      ~       Tl:i is LieeRse ReRewal UFS/\R S1:11313lemeRt, as revises 13er LieeRse GoRElitioR 12(a) al3ove, Eleseril3es eertaiR 13ro§rams to 13e im13lemeRtea aRel aeti*w<ities to 13e eom13 leteel 13rior to tl:ie 13erioel of m<teReleel 013eratioR .
              +.          Tl:ie lieeRsee sl:iall im13lemeRt tl:iose Rew 13ro§rams aRel eRRaReemeRts to e)<istiR§ J3FO§rams RO later tl:iaR May e,
                          ~
              ~           Tl:ie lieeRsee sl:iall eom13lete tl:iose aetivities as Roteel iR tl:ie GemmitmeRts a1313 li eal3le to ByroR URit 2 iR tl:iis S1:11313 lemeRt Ro later tl:iaR May e, 202e , er tl:ie eRel of tl:ie last ref1:1eli R§ 01;1ta§e 13rior to tl:ie 13erioel of m<teReleel 013eratioR , wl:iisl:iever oes1:1rs later.

d-,. Tl:ie lieeRsee sl:iall Ratify tl:ie ~JRG iR writiR§ witl:iiR d0 elays after l:iaviR§ aseom13lisl:ieel item (13)1 al3ove aRel iRsl1;1ele tl:ie stat1:1s of tl:iose astivities tl:iat l:iave 13eeR or remaiR to 13e eem13 letea iR item (13)2 al3eve. (13) Adoption of 10 CFR 50.69. "Risk-informed categorization and treatment of structures. systems. and components for nuclear power plants" Exelon is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding , and internal fire ; the shutdown safety assessment process to assess Renewed License No. NPF-66 Amendment No . ~

shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in the license amendment No. 204, dated October 22, 2018. Exelon will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process . Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e .g., change from a INSERT: seismic margins approach to a seismic probabilistic risk assessment LICENSE approach). CONDITION 2.C.(14) D. The facility requires no exemptions from the requirements of 10 CFR Part 50. An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1916, issued March 4, 1985, and relieved the licensee from the requirement of having a criticality alarm system. Therefore, the licensee is exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license. E. nie lioeRsee sl=lall i1T113le1TteRt aRel R=taiRtaiR iR effeot all 13Fovisi0Rs of tl=le a1313mveel

        .---------. 1' fire 13mteotioR J3FO§FOR=t as elesoFiseel iR tl=le lioeRsee 's Fire PmteotioR Re13ort aRel
        ~ tl=le lioeRsee 's le1teFS elateel ee13te1Ttl3eF 2d , 198e, OotoseF 2d , 198e, ~JO>v'eR=tser d ,

198e, DeoeR=teer 12 a Rel 19, 198e, a Rel daR1:10Ff 21 , 1987, aREl as a1313roveel iA tl=le eER elateel Fe8r1:1ary 1982 tl=IFOl:l§R e1:11313le1TteAt ~Jo . 8, s1:18jeot to tl=le followiA§ 13FovisioA: Tl::ie lioeRsee R=tay ITIOl(e 6R8R§8S to tRe a1313roveEI fire 13FoteetioR 13ro§r81Tt witl=lo1:1t 13rior a1313ro>v'al of tl=le GoR=tR=tissioA oRly if tl=lose ol=laA§es wo1:1lel Rot aelversely affeot tl=le asility to aol=lieve aRel R=taiRtaiR safe sl=l1:1telowR iR tl=le e>v'eAt of a fire . F. Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualifications, and safeguards contingency plans including amendments made Renewed License No. NPF-66 Amendment No. 2-14

pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50 .54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

                  "Byron Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 3," submitted by letter dated May 17, 2006.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 175 and modified by License Amendment No. 191 . G. Deleted H. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims . I. This renewed license is effective as of the date of issuance and sl=lall e ~(J3iFe at R=tielRi§Rt OR ~JoveR=tl3eF e, 2040 . FOR THE NUCLEAR REGULATORY COMMISSION is effective until the Commission notifies the licensee in writing that the IRA/ license is terminated. William M. Dean, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A-Technical Specifications (NUREG-1113)

2. Appendix B - Environmental Protection Plan

3. Appendix C - AEIElitioRal GoRElitioRs ~

Date of Issuance: November 19, 2015 1 The training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Renewed License No. NPF-66 Amendment No . ~

INSERT UNIT 2 RFOL LICENSE CONDITION 2.C.(14) ( 14) Handling of irradiated fuel in the spent fuel pool will not be permitted following implementation of the POTS until a minimum of 70 days following permanent shutdown.

ATTACHMENT 3 Markup of Technical Specifications Pages Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 MARKED UP TECHNICAL SPECIFICATIONS PAGES TS 1.1: Pages 1-9, INSERTS TS 1.2: Pages 1-3 TS1 .3: Pages1-14 TS 1.4: Pages 1-7 TS 3.0: Pages 1-6 TS 3.7.14: Page 1 TS 3.7.15: Pages 1-2 TS 3.7.16: Pages 1-3 TS 4.0: Pages 1-2 TS 5.1: Page 1 TS 5.2: Pages 1-2 TS 5.3: Page 1 TS 5.4: Page 1 TS 5.5: Pages 1-25 TS 5.6: Pages 1-7

TS Markups incorporate changes from of Byron Administrative Controls LAR submitted on September 24, 2020 Definitions 1.1 (NRC Accession No. ML20269A401) 1.0 USE AND APPLICATION 1.1 Definitions

        -------------------------------------NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Speci fication that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

        /\CTUATION LOGIC TEST          An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic CERTIFIED FUEL                       outpLit. The ACTUATimJ LOGIC TEST , as a minimLim ,
~HANDLER                                shall inclLide a continLiit~/ check of OLitpLit devices .
        /\XIAL FLUX DIFFERENCE         /\FD shall be the difference in norma l ized flLix CAFD)                          signals between the top and bottom halves of a t\m section excore neutron detector .

\V ... CALrnRATimJ /\ Cl=IMJ~JEL CALIBRATION shall be the adjListment , as

Cl=IMJ~JEL necessary , of the channel so that it responds

                                       ',Ji thin the reqLii red range and acm racy to knmm i npLits . The Cl=IMJ~JEL CALIBRATimJ sha l l encompass the entire channel , inclLiding the required sensor ,

alarm , interlock , display , and trip fLinctions . Calibration of i nstrnment channels .Ji th Resistance 1 INSERT 1 TemperatLire Detector (RTD) or thermocoLiple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjListable devices in the channel . +J::ie-Cl=IA~JNEL C/\LIBRATIO~J may be performed by means of any series of sequential , overlapping calibrations or total channel steps so that the entire channel is calibrated , and each step mList be performed

                                       ',Ji thin the FreqLiency in the SLirvei 11 ance FreqLiency Control Program for the devices included in the step .

BYRON - UNITS 1 &2 1.1 - 1 Amendment ~

Definitions 1.1 1.1 Definitions CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessFRent , b:Y observation , of channel behavior during operation . This deterFRination shall include , where possible , coFRparison of the channel indication and status to other indications or status derived froFR independent instruFRent channels FReasuring the saFRe paraFReter . CHA~l~l tl OPtRATIOW\L A COT shall be the injection of a siFRulated or TEST (COT) actual signal into the channel as close to the sensor as practicable to verif:Y the OPt:RABILITY of required alarm , interlock , displa:Y , and trip functions . The COT shall include adjustFRents , as necessary , of the required alarm , interlock , and trip setpoints so that the setpoints are within the required range and accuracy . The COT may be perforFRed by means of any series of sequential , overlapping , or total channel steps , and each step must be performed *,vi thin the Frequency in the Surveillance Frequency Control PrograFR for the devices included in the step . CORt ALTtRATIO~l CORE ALTt:RATiml shall be the FRoveFRent of any fb1el , sources , or reactivity control coFRponents , *11ithi n the reactor vessel *,,ii th the vessel head reFRoved and fuel in the vessel . Suspension of CORt ALTt:RATIONS shall not preclude coFRpletion of moveFRent of a component to a safe position . CORE OPERATHJG LIMITS The COLR is the unit specific docuFRent that REPORT (COLR) provides cycle specific paraFReter liFRits for the current reload cycle . These cycle specific paraFReter limits shall be determined for each reload cycle in accordance with Specifi ca ti on 5. 6. 5. Unit operation \Jithi n these liFRits is addressed in individual Specifications . BYRON - UNITS 1 &2 1.1 - 2 Amendment ~

Definitions 1.1 1.1 Definitions DOS E EQUIVALENT I 131 DOS E EQU IVALENT I 131 shall be that concentration of I 131 Cmicrocuries per gram) that alone would produce the same dose when inha l ed as the combined activities of iodine isotopes I 131 , I 132 , I 133 , I 134 , and I 135 actuall y present . +fie. determination of DOSE EQUIVALENT I 131 shall be performed using the Committed Effective Dose Equivalent (CEDE) 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 ." DOSE EQUIVALENT XE 133 DOSE EQUIVALENT XE 133 shall be that concentration of Xe 133 Cmicrocuries pe r 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 the minimum detectable activity . +fie. determination of DOS E EQU IVALENT XE 133 shall be performed using effective dose conversion factors for air submersion listed in Table III .1 of EPA Federa l Gui dance Report ~lo . 12 , 1993 , "External Exposure to Radionuclides in Air , Water , and Soil ." ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE ( ESF) RES PmlS E interval from when the monitored parameter ~ exceeds i t s ESF act uation setpoint at t he channel sensor until the ESF equipment is capable of performing its safety function Ci .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 , v. here appl i cable . +fie. 1 response time may be measured b:} means of any 1 se ries of sequential , overlapping , or total steps so that the entire response time is measured . Jn. l ieu of measurement , response time may be verified for selected components provided that the components and methodology for verifi cation have been previously rev i ei11ed and approved by the NRG . BYRON - UNITS 1 &2 1.1 - 3 Amendment -+/-B Definitions 1.1 1.1 Definitions IMSERVI CE TESTIMG The HJSERVI CE TESTI MG PROGRi\M is t he l i censee PROGR/\M program that fu l fills the requirements of 10 CFR 50 .55a(f) . LEAKAGE LEAKAGE shall be:

                  .a-. Identified LE/\K/\GE h     LE/\Kii\GE , such as t hat from pump seals or va l ve packing (except Reactor Cool ant pump (RC P) seal 't1ater injection or leakoff) ,

that is captured and conducted to collection systems or a sump or collecting tank;

                       .i..- LEAKii\GE into the containment atmosphere from sources that are both specifical ly l ocated and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKii\GE ; or J.- Reactor Cool ant System (RCS) LE/\KAGE through a Stearn Generator to the Secondary System (primary to secondary LEAKAGE);
                  -8.- Unidentified LEAKAGE All LEAKAGE (except RCP seal 'dater injection or l eakoff) that is not identified LEAKAGE; t.-   Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) throug h a noniso l abl e fault in an RCS component body , pipe wa l l , or vessel wa ll .

MASTER RELAY TEST A MASTER RELAY TEST shall consi st of energizing each master rel ay and verifying the OPERi\BI LITY of each relay . The M,i\STER RELAY TEST shall include a continuity check of each associated slave relay . BYRON - UNITS 1 &2 1.1 - 4 Amendment -+/--9+

TS Markups incorporate changes from of Byron Administrative Controls Definitions LAR submitted on September 24, 2020 1.1 (NRC Accession No. ML20269A401) 1.1 Definitions A MODE shall correspond to any one inclusive combination of core reactivity condition , power level , average reactor coolant temperature , and NON-CERTIFIED reactor vessel head closure bolt tensioning OPERATOR specified in Table 1. 1 1 .:ith fuel in the reactor 1 vessel . OPERABLE OPERABILITY A system , subsystem , train , component , or device shall be OPERABLE or have OPERABILITY \*then it is ca pa bl e of performing its specified safet:Y function(s) and ..:hen all necessary attendant 1 instrumentation , controls , normal or emergency INSERT 2 electrical power , cooling and seal water , lubrication , and other auxiliary equi pment 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) . PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation . These tests are:

                                       -a.-   Described in Chapter   1~ , Initial Test Program ,

of the UFSAR;

                                       -b.-   Authorized under the provi si ans of 10 GFR 50 .59; or
                                       -&.-   Otherwise approved by the    ~Juel ear Regulatory Commission .

BYRON - UNITS 1 &2 1.1 - 5 Amendment ~

Definitions 1.1 1.1 Definitions PRESSURE /\ND The PTLR is the unit specific document that TEMPER~TURE LIMITS provides the reactor vessel pressure and REPORT CPTLR) temperature limits including heatup and cooldown rates , and the pressurizer Power Operated Relief Valve CPORV) lift settings for the current reactor vessel fluence period . These pressure and temperature limits shal l be determined for each fl uence period in accordance \Ji th Specification 5.6.6. QU/\DM~IT PO\\'ER TI LT QPTR shall be the ratio of the maximum upper MTIO CQPTR) excore detector calibrated output to the average of the upper excore detector calibrated outputs , or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outp~ts , whichever is greater . MTED T~ERM/\L P01..JER RTP shall be a total reactor core heat transfer CRTP) rate to the reactor coolant of 3645 MWt . REACTOR TRIP The RTS RESPO~J S E TIME shall be that time interval SYSTEM CRTS) RESPOMS E from 1:1hen 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 b:J' 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 pre vi ousl y revi mmd and approved by the MRC . REC EMT LY I RMDIATED rnEL ~uel that has occupied part of a critical reactor core *,Jithi n the previous 48 hours . ~l ote that all fuel that has been in a critical reactor core is referred to as irradiated fuel . BYRON - UNITS 1 &2 1.1 - 6 Amendment -+/-fil

Definitions 1.1 1.1 Definitions SHUTD0 1/M M/\RGHJ (SOM) 1 SOM shall be the instantaneous amount of reactivity by v1hi ch the reactor is subcriti cal or would be subcritical from its present condition assuming:

                        -a-. All Rod Cluster Control Assemblies CRCCAs) are fully inserted except for the single RCCA of highest reactivity *,1orth , *,1hi ch is assumed to be fully withdrawn . With any RCCA not capable of being fully inserted , the reactivity *.mrth of the RCC/\ must be accounted for in the determination of SOM ; and ft.. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperature .

SLAVE RELAY TEST /\SLAVE REL/\Y TEST shall consist of energizing each slave relay and verifying the OPER/\BI LITY of each slave relay . The SLAVE RELAY TEST shall include , as a minimum , a continuity check of associated testable actuation devices . STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems , subsystems , channels , or other designated components during the interval specified by the Survei l lance Frequency , so that all systems , subsystems , channels , or other designated components are tested during n Surveillance Frequency intervals , where n is the total number of systems , subsystems , channels , or other designated components in the associated function . THERMAL PO'..JER THERMAL POV. ER shall be the total reactor core heat 1 transfer rate to the reactor coolant . BYRON - UNITS 1 &2 1.1 - 7 Amendment ~

Definitions 1.1 1.1 Definitions TRIP ACTU/\THJG DEVICE /\ TADOT shall consist of operating the trip OPER/\TIOW\L TEST actuating device and verifying the OPERl\BI LITY of CT/\DOT) required alarm , interlock , display , and trip functions . The TADOT shall include adjustment , as necessary , of the trip actuating dev i ce so that it actuates at the required setpoint within the required accuracy . The TADOT may be performed by means of any series of sequential , overlapping , or total channel steps , and each step must be performed '.Ji thin the Frequency in the Survei 11 ance Frequency Control Program for the devices included in the step . BYRON - UNITS 1 &2 1.1 - 8 Amendment ~

Definitions 1.1 Tebl e 1. 1 1 (pege 1 of 1) MODES

                                                            ?£ Ril\TEQ         NlERl\GE REAGT +/-\J +/-T¥     Tl=lERMAb     REAGTOR GOObAMT MOOf.                 THbE               Gmmnrn ~i         Pm~tR/M        TEMPE Ril\TbJ RE
                                             ~                                   ~
    -+/--        Pm1er Opereti on             ~~                 >  .§.             AA
    ~         Stertup                       ~~                 ~ .§.              AA J         l=lot Stendby                 4 ~                 AA              ~JW 4         ~     Shutdmm (hl             4~                  AA        JW>~> ~
    .§.       W+4 Shutdmm(hl                4 ~                 AA              ~~
    .e.        Refuel i n~t8                  AA                AA                AA W        Excluding decey heet .

+.9+ All r equi r ed r ee ct or vess el heed cl os ur e bolts f61lly ten si oned . +G+ One or more required ree ctor vessel heed closure bolts less then fully ten si oned . BYRON - UNITS 1 &2 1.1 - 9 Amendment ~

TS 1.1 INSERTS INSERT 1 A CERTIFIED FUEL HANDLER is an individual who complies with the provisions of the CERTIFIED FUEL HANDLER training and retraining program required by Specification 5.3.2. INSERT 2 A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1 but is not a CERTIFIED FUEL HANDLER.

Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors. Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions , Required Actions, Coffipletion Tiffies , Survei 11 ances , and Frequencies . The only l ogi cal connector-& that appear in TS trre AND and .QB . The physical arrangement of these c ectors ~ n 2.!2..tutes logical convent ions with specific meanings. ~ BACKGROUND Several l evels of logic may be used to state Required r11 Actions.1' These levels are identified by the placement (or

            ~ of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.
                \JAen lo§ical connectors are used to state a Condition ,

Coffipl eti on Tiffie , Surveillance , or Freq1:Aency , only Hie first level of lo§ic is 1:Ased , and tAe lo§ical connect or is left justified \JitA tAe stateffient of tAe Condition , Coffipletion Tiffie , S1:Arvei 11 ance , or Freq1:Aency . EXAMPLE& The following exampl e-& illus~t7ri the use of logical connectors. ~ BYRON - UNITS 1 &2 1.2 - 1 Amendment -+/-Ge-

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLE.£ (cont inued) EXAMPLE 1. 2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify . . . AND A.2 Restore . . . In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed. BYRON - UNITS 1 &2 1.2 - 2 Amendment -+/-Ge-

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES ( COAt i Al::led) EX#4PLE 1. 2 2 ACTIONS CONDIT Imi REQUIRED ACTION cm4PLETimi TH4E I\ M. LCO AOt Fflet . A-:+ Trip GR A.2. 1 Veri f:y MID A.2.2.1 Red1::1ce GR A.2.2.2 PerforFFl GR

                                                 ~         A~~§A Hii s eHaFFlpl e represeAts a Fflore coFFlpl i cated 1::1se of 1ogi cal coAAectors . Req1::1ired P.ctioAs A.1, A.2, aAd P. .3 are alternative choices , oAl y oAe of 1Jhi ch Ffll::lst be perforFFled as 1

i Adi cated by Hie 1::1se of Hie 1ogi cal COAAector QB aAd the left j1::1stified plaCCFFlCAt . AA)' SAC of these three ActiOAS FF1ay be choseA . If A.2 is choseA , theA both P. .2.1 aAd A.2.2 FF11::1st be perforFF1ed as i Adi cated by the 1ogi cal connector filill . Req1::1ired ActioA A.2.2 is FF1et by perforFF1iAg A.2.2.1 or A.2.2.2. The iAdented positioA of the logical coAAector QB iAdicates that A.2.2.1 aAd A.2.2.2 are alternative choices , only one of v1hi ch FF11::1st be perforFFled . 1 BYRON - UNITS 1 &2 1.2 - 3 Amendment -+/-Ge-

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. storage and handling of irradiated fuel BACKGROUND Limiting Conditions for Opera n CLCOs) specify minimum requirements for ensuring safe operation of the unit . The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each st ated Condition are Required Action(s) and Completion Time(s). DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the

                -w:H-t is in a MODE or specified condition stated in the Applicability of the LCO. Unless otheP11'i se specified , the Completion Ti me begins ',Jhen a senior licensed operator on the operating shift crew with responsibility for pl ant operations makes the determination that an LCO is not met and an ACTimlS Condition is entered . The "othen1i se specified " exceptions are varied , such as a Required /\ction
                 ~l ate or Survei 11 ance Reqbli rement Note that provides an alternative time to perform specific tasks , such as testing ,
                \Jithoblt starting the Completion Time . While blt ilizing the Note , shoblld a Condition be applicable for any reason not addressed by the Note , the Completion Time begins . Should the time allo\rance in the Note be exceeded , the Completion Time begins at that point . The exceptions may also be incorporated into the Completion Time . for example , LCO 3J:L1 , "AC Soblrces Operating ," Reqbli red /\cti on g. 3 ,

reqbli res declaring reqbli red featblre( s) Sblpported b:J' an inoperable diesel generator , inoperable Hhen the redundant reqblired feature(s) are inoperable . The Completion Time states , "4 hours from discovery of Condition g concurrent

                ',Ji th i noperabi l i ty of redundant required feature( s) ." -+/--n-thi s case the Completion Time does not begin blntil the conditions in the Completion Time are satisfied . Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the -w:H-t is not within the LCO Applicability.                 ~

If situations are discovered that require entry into more BYRON - UNITS 1 &2 1.3 - 1 Amendment ~

Completion Times 1.3 1.3 Completion Times DESCRIPTION (continued) than one Condition at a ti me ',Jithi n a single LCO (mbllti pl e Conditions) , the Reqblired Actions for each Condition mblst be performed within the associated Completion Time . When in mblltiple Conditions , separate Completion Times are tracked for each Condition starting from the discovery of the sitblation that reqblired entry into the Condition , blnless othenJi se specified . Once a Condition has been entered , sblbseqblent t rains , Sblbsystems , components , or variables expressed in the Condi ti on , discovered to be inoperable or not .1ithi n limits , 1

                '11 i 11 .BQ1 resb1 1

l t in separate entry into the Condition , blnl ess specifically stated . The Reqblired Actions of t he Condition continble to apply to each additional failure , with Completion Times based on initial entry into the Condition , blnless otherwise specified . However , when a sblbsegblent train , sblbsystem , component , or variable expressed in the Condition is discove red to be inoperable or not within limits , the Completion Time(s) may be extended . To apply this Completion Ti me ext ension , t .:o 1 cri t eria mblst f i rst be met . The Sblbseqblent i nope rability:

                .fr.      Mblst exist conrnrrent 'di th the first i noperabi l i ty ;
                          -aHG                              ~-
                -b.-      Mblst remain inoperable or not ',Jithin limits after the first i noperabil ity is resolved .

The total Completion Time all m1ed fo r completing a Reqbli red Action to address the Sblbseqblent inoperability shall be limited to the more restrictive of either: tt. The stated Completion Time , as measblred from the initial entry into the Condition , plbls an additional 24 hoblrs; or

                -b.-      The stated Completion Time as measblred from discovery of the Sblbseqblent inoperability .

The above Completion Time extension does not apply to those

                 ~pecifications that have exceptions that allm: completely separate re entry into the Condition (for each train ,

subsystem , component , or variable expressed in the Condition) and separate tracking of Completion Times based on this re entry . These exceptions are stated in individblal

                 ~pecifications .

BYRON - UNITS 1 &2 1.3 - 2 Amendment ~

Completion Times 1.3 1.3 Completion Times DESCRIPTION (continued) The above Completion Time extension does not apply to a Completion Time .:ith a modified "time zero ." This modified 1

                 "time zero " may be expressed as a repetitive time Ci .e., "once per 8 hours ," ',Jhere the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified b:J' the phrase "from discovery . . . " c--{Ilt-- - - - - - - -

EXAMPLE~ The following exampl e-& illustrate i{he use of Completion Times with different types of Conditions and changing Condi ti~ns .

                            ,__---ilRequired Actions I EXAMPLE 1. 3-1 ACTIONS CONDITION             REQUIRED ACTION    COMPLETION TIME B. Required            8.1 ge in MOQE 3. 9 hOblrS Action and associated        AND    ~                  ~

Completion Time not 8.2 ge in MOQE 5. 39 hours met.

                                                 ~                  '--i1mmediately I Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered.

The Required Actions of Condition g are to be in MOQE 3

                ',,'ithi n 9 hours Mm in MOQE 5 \Jithi n 39 hours . A total of 9 hours is allowed for reaching MOQE 3 and a total of 36 hours (not 42 hours) is allowed for reaching MODE 5 from the ti me that Condition g .:as entered . If MODE 3 is reached 1

within 3 hours , the time allowed for reaching MOQE 5 is the next 33 hours because the total time allo\rod for reaching MODE 5 is 39 hours . If Condition g is entered \i'hile in MODE 3, the time allmrod for reaching MODE 5 i s the next 39 hours . BYRON - UNITS 1 &2 1.3 - 3 Amendment ~

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued) EXAMPLE 1. 3 2 ACTIOMS co~mnrn~i REQbJ mm ,Cl,CT ION COMPLETION TIME nA . One pump A.+/-- Resto Fe pump to 7 days inopeFable . OPER/\BLE status .

                    ~L  RequiFed       -lhJ-   Be in MOQE 3.       9 ROUFS Action and associated     AfiID Completion Time not       ~       Be in MOQE ~ .      39 ROUFS met.-
                \~hen    a pump is declaFed inopeFable , Condition A is enteFed .

If the pump is not FestoFed to OPER/\BLE status within 7 days , Condition g is also enteFed and the Completion Time clocks foF RequiFed Actions g.1 and B.2 staFt . If the inopeFable pump is FestoFed to OPERABLE status afteF Condition B is enteFed , Condition A and g aFe exited , and theFefoFe , the RequiFed Actions of Condition B may be teFminated . v.'hen a second pump is decl aFed i nopeFabl e .Jhil e the fiFst 1 pump is still inopeFable , Condition A is not Fe enteFed foF t he second pump . LCO 3.0.3 is enteFed , since t he ACTIONS do not include a Condition foF moFe than one inopeFable pump . The Compl etion Time cl ock foF Condition A does not stop afteF LCO 3.0.3 is enteFed , but continues to be tFacked from the time Condition A ',Jas initially enteFed . While in LCO 3.0.3, if one of the inopeFable pumps is FestoFed to OPERABLE status and the Completion Time foF Condition A has not expiFed , LCO 3.0.3 may be exited and ope Fa ti on continued in accoFdance ',Ji th Condi ti on A. BYRON - UNITS 1 &2 1.3 - 4 Amendment ~

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued) While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time fo r Condition A has expired , LCO 3.0.3 may be exited and operation continued in accordance \Ji th Condi ti on B. -Tfte. Compl etion Time for Condition B is tracked from the time the Condition A Compl etion Time expired . On restoring one of the pumps to OP E R~ B L E status , t he Condition A Completion Time is not reset , but continues from the time the first pump 11as declared inoperable . -Tl+i-& 1 Completi on Time may be extended i f the p~mp rest ored to OPERABLE stat~s was the first inope rable p~mp . ;~ 24 hobir extension to the stated 7 days is allowed , provided this does not res~lt in the second p~mp being inoperable for

                > 7 days .

BYRON - UNITS 1 &2 1.3 - 5 Amendment ~}}

v t e Letter Sequence Request
EPID:L-2020-LLA-0261, TSTF-372 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement Administration Questions 14 April 2021 21 June 2022 Answers 20 July 2022 Results Approval Other: ML21264A672, ML23188A051
MONTHYEAR2021-04-14 [Table View]

ML20303A326

Text

3.0 REGULATORY EVALUATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

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