RS-20-132, License Amendment Request, Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

License Amendment Request, Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

ML20303A313
Person / Time
Site:	Dresden
Issue date:	10/29/2020
From:	Simpson P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-20-132
Download: ML20303A313 (201)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-20-132 10 CFR 50.90 October 29, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2, and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

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 Dresden Nuclear Power Station, Units 2 and 3,"

dated September 2, 2020 (NRC Accession No. ML2024627)

2. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Request for Approval of Certified Fuel Handler Training 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 NRC, "License Amendment Request - Proposed Changes to Unit 1 Technical Specifications Section 6.1, 'Responsibility,' and Units 2 and 3 Technical Specifications 1.1, 'Definitions,' and 5.0, 'Administrative Controls,' for Permanently Defueled Condition," dated September 24, 2020 (NRC Accession No. ML20269A404)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests amendments to the Renewed Facility Operating Licenses (RFOLs) and Appendix A, Technical Specifications (TS),

of RFOL Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station, Units 2 and 3 (Dresden). 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 Dresden Permanently Defueled Technical Specifications (PDTS).

By letter dated September 2, 2020 (Reference 1), Exelon provided formal notification to the U.S. Nuclear Regulatory Commission (NRC) of Exelon's determination to permanently cease operations at Dresden on or before November 30, 2021. Once the certifications for permanent

October 29, 2020 U.S. Nuclear Regulatory Commission Page 2 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)(i) and (ii), 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. In support of this condition, the Dresden RFOLs and associated TS are being proposed for revision to reflect the planned permanent shutdown and defueled condition in accordance with 10 CFR 50.36(c)(6), "Decommissioning."

The bases for the proposed amendments are that certain license conditions, TS requirements, and current licensing basis may be revised or removed to reflect the permanently defueled condition. In general, the changes propose the elimination of those TS applicable in operating conditions where fuel is placed in the reactor vessels. Changes to other TS limiting conditions for operation, definitions, surveillance requirements, administrative controls, and license conditions are also proposed. The proposed amendments would modify the 10 CFR 50 licenses and the TS to make those changes.

The NRC is currently reviewing the Certified Fuel Handler Training and Retraining Program that was submitted on September 24, 2020 (Reference 2). In addition, the NRC is currently reviewing proposed changes to the organization, staffing, and training requirements contained in Section 6.1, "Responsibility," of the Dresden, Unit 1, TS and Section 5.0, "Administrative Controls," of the Dresden, Units 2 and 3, TS that were submitted on September 24, 2020 (Reference 3). These two referenced licensing actions complement and support this proposed license amendment request. to this letter provides a detailed description and evaluation of the proposed changes. Attachments 2 and 3 provide the existing Dresden RFOL and TS pages, respectively, marked up to show the proposed changes. TS sections that are deleted in their entirety are identified as such, but the associated deleted pages are not included in Attachment 3.

Proposed changes to the TS Bases addressing the proposed changes to the remaining TS are provided for information in Attachment 4.

The proposed changes have been reviewed by the Dresden Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendments by October 29, 2021, to support the current schedule for the Dresden transition to a permanently defueled facility. Exelon requests that the approved amendments become effective following the submittal of the required 10 CFR 50.82(a)(1)(ii) certification that Dresden, Units 2 and 3, have been permanently defueled. Once approved, the amendments shall be implemented within 60 days from the effective date of the amendments. As discussed in Attachment 1, a 48-day decay period is needed to meet the dose limits for the Fuel Handling Accident. In order to implement the PDTS within 60 days following permanent defueling, Exelon is requesting two license conditions, one for each unit, as proposed in Attachment 1, that restrict handling of irradiated fuel in the spent fuel pools during that 48-day time period following permanent shutdown.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b),

Exelon is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Officials.

October 29, 2020 U.S. Nuclear Regulatory Commission Page 3 There are no regulatory commitments contained within this submittal.

Should you have any questions concerning this submittal, please contact Mr. Mitchel Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of October 2020.

Respectfully, Patrick R. Simpson Sr. Manager - Licensing Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Markup of Renewed Facility Operating License Nos. DPR-19 and DPR-25

3. Markup of Technical Specifications Pages

4. Markup of Technical Specifications Bases Pages (For Information Only) cc: w/ Attachments NRC Regional Administrator, Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Project Manager, NRR - Dresden Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

Attachment 1 Evaluation of Proposed Changes

Subject:

Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Conditions 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements/Criteria 3.2 No Significant Hazards Consideration 3.3 Conclusion

5.0 REFERENCES

Attachment 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon), requests amendments to the Renewed Facility Operating Licenses (RFOLs) and Technical Specifications (TS), of RFOL Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station, Units 2 and 3 (Dresden). 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 Dresden, Units 2 and 3, Permanently Defueled Technical Specifications (PDTS).

By letter dated September 2, 2020 (Reference 1), Exelon provided formal notification to the U.S.

Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.4(b)(8) and 10 CFR 50.82(a)(1)(i) that it would permanently cease operations at Dresden on or before November 30, 2021. Once the certifications for permanent cessation of operations and removal of fuel from the reactor vessels are submitted to the NRC for Dresden, Units 2 and 3, pursuant to 10 CFR 50.82(a)(1)(i) and (ii), NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the Units 2 and 3, reactors or emplacement or retention of fuel in the reactor vessels under the 10 CFR 50 licenses. In support of this condition, the Dresden Unit 2 and Unit 3 RFOLs and associated TS are being proposed for revision to reflect the planned permanent shutdown and defueled condition in accordance with 10 CFR 50.36(c)(6),

"Decommissioning."

The proposed changes to the TS and RFOLs are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes also include editorial changes to format (i.e., margin, font, tabs, etc.) of content; revised numbering of sections and pages, and the deletion of unused placeholders, where appropriate, to condense the number of pages in the TS without affecting the technical content.

The current Dresden TS contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe operation of the facility, including the safe storage and handling of irradiated fuel. Since the safety function related to safe storage and handling of irradiated fuel at an operating plant is similar to the corresponding function at a permanently defueled facility, the existing TS provide an appropriate level of control. However, the majority of the existing TS are only applicable with the reactors in an operational mode. LCOs and associated Surveillance Requirements (SRs) that will not apply in the permanently defueled condition are being proposed for deletion. The remaining portions of the TS are being proposed for revision and incorporation as the PDTS to provide a continuing acceptable level of safety that addresses the reduced scope of postulated design basis accidents associated with a permanently defueled facility.

The proposed changes would become effective following submittal of the certifications required by 10 CFR 50.82(a)(1)(i) and (ii). As discussed below, a 48-day decay period is needed to meet the dose limit for the Fuel Handling Accident (FHA). In order to implement the PDTS following permanent defueling, Exelon is requesting two license conditions, one for each unit, as proposed below, that restrict irradiated fuel handling during that 48-day time period following permanent shutdown.

Page 1 of 85

Attachment 1 Evaluation of Proposed Changes Related Licensing Actions By letter dated September 24, 2020 (Reference 2), Exelon submitted a Certified Fuel Handler (CFH) Training and Retraining Program for NRC approval. By letter dated September 24, 2020 (Reference 3), Exelon submitted a license amendment request (LAR) to the NRC that proposed changes to the organization, staffing, and training requirements in Dresden TS Section 6.1, "Responsibility," of the Unit 1, TS, and Section 1.1, "Definitions," and Section 5.0, "Administrative Controls," of the Units 2 and 3, TS, which are incorporated into this LAR. The CFH program and the referenced LAR will become effective and will be implemented once the Dresden, Units 2 and 3, reactors have been defueled and the certifications of permanent removal of fuel from the reactor vessels have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1)(ii). These licensing actions complement and support this proposed LAR.

2.0 DETAILED DESCRIPTION The proposed amendments would modify the Dresden RFOLs and revise the operating TS into the Dresden PDTS to comport with a permanently defueled condition, as well as clarifying current licensing bases to reflect the permanently defueled condition. To support the proposed changes, Exelon has evaluated the Design Basis Accidents (DBAs) that will be applicable in a permanently shutdown and defueled condition. The DBA evaluation provides the framework for the proposed changes.

Design Basis Accident Analyses Applicable to Proposed Change Chapter 15 of the Dresden Updated Final Safety Analysis Report (UFSAR) contains the DBAs and transient scenarios applicable to Dresden, Units 2 and 3, including during power operations.

The most severe postulated accident for a nuclear power plant involves damage to the nuclear reactor core and the release of large quantities of fission products to the reactor coolant system (RCS). Many of the accident scenarios postulated in the UFSAR involve failures or malfunctions of systems that could affect a reactor core.

With the termination of reactor operations at Dresden and the permanent removal of fuel from the Units 2 and 3, reactor vessels as certified in accordance with 10 CFR 50.82(a)(1 )(i) and (ii),

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), the majority of the DBA scenarios postulated in the UFSAR will no longer be possible. During decommissioning, the irradiated fuel will be stored in the spent fuel pools (SFPs) or in the Independent Spent Fuel Storage Installation (ISFSI) until it is shipped offsite in accordance with the schedules to be provided in the Post Shutdown Decommissioning Activities Report (PSDAR) and the Spent Fuel Management Plan. With the reactors permanently shutdown and defueled, many associated systems and instrumentation, as well as the turbine generator, will no longer be in operation and have no function related to the safe storage and handling of irradiated fuel.

Chapter 15 of the Dresden UFSAR describes the safety analyses that were evaluated to demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed regulatory requirements. Two basic groups of events are pertinent to safety, which are abnormal operational transients and postulated DBAs; these two Page 2 of 85

Attachment 1 Evaluation of Proposed Changes groups were investigated separately. The analyses of the abnormal operational transients evaluate the ability of the plant protection features to ensure that, during these transients, no fuel damage occurs, and the RCS pressure limit is not exceeded. The safety design limits require that damage to the fuel be limited and that no nuclear system process barrier damage results from any abnormal operational occurrence. Thus, analysis of this group of events evaluates the features that protect the first two radioactive material barriers. Analysis of the events in the second group, postulated DBAs, evaluates situations that require functioning of the engineered safeguards in order to protect the fission product barriers, including containment, in order to minimize the offsite radiological consequences.

Safety analyses are integral of the plant's design and licensing basis. The safety analyses demonstrate the integrity of the fission product barriers, the capability to shutdown the reactors and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents and transients. Systems, structures, and components (SSCs) that perform design basis functions are credited in the safety analyses for the purpose of mitigating the transient or accident.

The potential accidents or transients have been re-evaluated for the permanently defueled condition where all fuel has been removed from the reactors and placed in the SFPs. A list of Chapter 15 events and whether each applies to a permanently defueled condition is provided in Table 2.1 below. The UFSAR Chapter 15 events that remain applicable to Dresden in the permanently shutdown and defueled condition, with fuel stored in the SFPs are: 1) a postulated liquid releases due to liquid tank failure, and, 2) a fuel handling accident (FHA) in the SFP (which is a new analysis, since the existing FHA in Chapter 15 addresses an FHA in the reactor core).

Table 2.1 UFSAR Permanently Defueled Postulated Accident or Transient Section Applicability Increase in Heat Removal by the Reactor 15.1 Coolant System Decrease in Feedwater Temperature (Loss of 15.1.1 Not Applicable Feedwater Heating)

Increase in Feedwater Flow (Feedwater 15.1.2 Not Applicable Controller Malfunction) - Maximum Flow 15.1.3 Increase in Steam Flow Not Applicable Decrease in Heat Removal by the Reactor 15.2 Coolant System 15.2.1 Steam Pressure Regulator Malfunction Not Applicable 15.2.2.1 Generator Load Rejection Without Bypass Not Applicable Generator Load Rejection With Bypass 15.2.2.2 Not Applicable System (Loss Of Electrical Load) 15.2.3.1 Turbine Trip Without Bypass Not Applicable Turbine Trip With Partial Bypass - Maximum 15.2.3.2 Not Applicable Power Inadvertent Closure of Main Steam Line 15.2.4 Not Applicable Isolation Valves Page 3 of 85

Attachment 1 Evaluation of Proposed Changes UFSAR Permanently Defueled Postulated Accident or Transient Section Applicability 15.2.5 Loss of Main Condenser Vacuum Not Applicable 15.2.6 Loss of Offsite AC Power Not Applicable Loss of Normal Feedwater Flow (Feedwater 15.2.7 Not Applicable Controller Malfunction) - Zero Flow 15.2.8 Loss of Stator Cooling Not Applicable 15.3 Decrease in Reactor Coolant System Flowrate 15.3.1 Single and Multiple Recirculation Pump Trips Not Applicable Recirculation Flow Controller Failure 15.3.2 Not Applicable (Malfunction) - Decreasing Flow 15.3.3 Recirculation Pump Shaft Seizure Not Applicable Recirculation Pump Shaft Seizure While in 15.3.4 Not Applicable Single Loop Operation 15.3.5 Recirculation Pump Shaft Break Not Applicable 15.3.6 Jet Pump Malfunction Not Applicable 15.4 Reactivity and Power Distribution Anomalies Uncontrolled Control Rod Assembly 15.4.1 Not Applicable Withdrawal - Subcritical or Startup Condition 15.4.2 Rod Withdrawal Error Not Applicable 15.4.3 Control Rod Maloperation Not Applicable Startup of Idle Recirculation Loop at Incorrect 15.4.4 Not Applicable Temperature (Cold Recirculation Loop)

Recirculation Flow Controller Failure 15.4.5 Not Applicable (Malfunction) - Increasing Flow 15.4.7 Mislocated Fuel Assembly Not Applicable 15.4.8 Misoriented Fuel Assembly Not Applicable 15.4.10 Control Rod Drop Not Applicable 15.4.11 Thermal Hydraulic Instability Not Applicable 15.5 Increase in Reactor Coolant Inventory Inadvertent Actuation of High Pressure 15.5.1 Not Applicable Coolant Injection During Power Operation 15.6 Decrease in Reactor Coolant Inventory Inadvertent Opening of a Safety Valve, Relief 15.6.1 Not Applicable Valve, Or Safety Relief Valve Break in Reactor Coolant Pressure Boundary 15.6.2 Not Applicable Instrument Line Outside of Containment Steam System Line Break Outside the 15.6.4 Not Applicable Containment Loss-of-Coolant Accidents Resulting from 15.6.5 Not Applicable Piping Breaks Inside Containment Radioactive Release from a Subsystem or 15.7 Component Radioactive Gas Waste System Leak or 15.7.1 Not Applicable Failure Postulated Liquid Releases Due to Liquid Tank 15.7.2 Applicable Failure Page 4 of 85

Attachment 1 Evaluation of Proposed Changes UFSAR Permanently Defueled Postulated Accident or Transient Section Applicability Design Basis Fuel Handling Accidents During 15.7.3 Applicable for FHA in SFP Refueling 15.8 Anticipated Transients Without Scram 15.8.1 Closure of Main Steam Isolation Valves Not Applicable 15.8.2 Loss of Normal AC Power Not Applicable 15.8.3 Loss of Normal Feedwater Flow Not Applicable 15.8.4 Turbine-Generator Trip Not Applicable 15.8.5 Loss of Condenser Vacuum Not Applicable 15.8.6 Increased Steam Flow Evaluation Not Applicable Pressure Regulatory Failure - Open to 15.8.7 Not Applicable Maximum Demand Fuel Handling Accident Analysis for the Permanently Defueled Condition The FHA event currently provided in UFSAR Section 15.7.3 pertains to postulated mechanical damage to the fuel rod cladding from a fuel assembly being dropped on the reactor core, and bounds an FHA in the SFP. Once the fuel has been permanently removed from the reactors, this accident will no longer be credible. Therefore, new FHA analysis was performed to determine the Control Room, Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses at Dresden due to an FHA over an SFP post-cessation of power operations, after all fuel has been removed from the reactor. This analysis conforms to the Regulatory Guide 1.183 (Reference 4) and RIS 2006-04 (Reference 5) methodology, and provides limiting doses for Dresden.

This analysis did not credit secondary containment, secondary containment isolation or filtration by the Standby Gas Treatment (SGT) System. The FHA analysis determined that, following a 48-day decay period, Control Room Emergency Ventilation (CREV) is not required to maintain dose consequences for control room occupants within the criteria of 10 CFR 50.67(b)(2)(iii).

Consequently, TS LCOs and SRs associated with CREV and support equipment are proposed for deletion. This FHA analysis concludes that a decay time of 48 days is required to meet the limits of 10 CFR 50.67 and Regulatory Guide 1.183.

The analysis was performed using conservative values for assumptions and inputs applicable to either of the Dresden units; therefore, the results are conservative for both units. As discussed below, two new license conditions, one for each unit, are proposed to restrict handling of irradiated fuel in the SFPs until 48 days have elapsed following the permanent shutdown of Dresden.

Detailed Review of General Design Criteria After Permanent Defueling Dresden was not licensed to the 10 CFR 50, Appendix A, General Design Criteria. Dresden UFSAR, Section 3.1.1 contains an evaluation of the design basis of Dresden with respect to the first draft of the 70 proposed "General Design Criteria for Nuclear Power Plant Construction Permits" issued by the Atomic Energy Commission in July 1967. The design basis of Unit 2, was later evaluated against the final "General Design Criteria for Nuclear Power Plants,"

Page 5 of 85

Attachment 1 Evaluation of Proposed Changes published as 10 CFR 50, Appendix A in July 1971. This evaluation is presented in Section 3.1.2 of the Dresden UFSAR.

Detailed Discussion of Proposed RFOL and TS Changes The following tables identify each RFOL and TS section that is being changed, the proposed change, and the basis for each change. Changes to the Unit 2, RFOL are addressed first followed by the Unit 3, RFOL, and then the TS. Proposed insertions are shown in Bold-Italics, and deletions are shown using strikethrough.

Attachments 2 and 3 provide the marked-up version of the Dresden, Unit 2 and Unit 3, RFOLs and TS, respectively. The TS that are deleted in their entirety are identified as such below, but the associated deleted pages are not included in Attachment 3. Proposed changes to the TS Bases addressing the proposed changes to the relevant TS are provided for information in . Upon approval of the amendments, changes to the Bases will be incorporated in accordance with TS 5.5.10, "Technical Specifications (TS) Bases Control Program," which will be retained in the PDTS with minor changes.

Additionally, the proposed changes to the TS are considered a major rewrite; therefore, revision bars are not used. Revised formatting (margins, font, tabs, etc.) of content is used to create a continuous electronic file, revised numbering of sections and pages; and the deletion of unused placeholders, where appropriate, is used to condense and reduce the number of pages in the TS without affecting the technical content.

10 CFR 50.36, "Technical specifications," promulgates the regulatory requirements related to the content of TS. As detailed in subsequent sections of this request, this regulation lists four criteria to define the scope of equipment and parameters that must be included in TS (see Section 3.1). In a permanently defueled condition, the scope of equipment and parameters that must be included in the Dresden, Units 2 and 3, PDTS is limited to those needed to address the remaining applicable DBAs so that the consequences of the accidents are maintained within acceptable limits.

The following tables identify each RFOL section that is being changed, the proposed change, and the basis for the change.

Detailed Description of the Proposed Change to the Dresden, Unit 2, RFOL Unit 2, RFOL Title Current Proposed Renewed Facility Operating License Renewed Facility Operating License Basis The License title is modified to eliminate the reference to "Operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Page 6 of 85

Attachment 1 Evaluation of Proposed Changes Finding 1.B Current Proposed Construction of the Dresden Nuclear Power Station, Deleted; Construction of the Dresden Nuclear Power Unit 2 (the facility) has been completed in conformity Station, Unit 2 (the facility) has been completed in with Construction Permit No. CPPR-18 and the conformity with Construction Permit No. CPPR-18 application, as amended, the provisions of the Act, and the application, as amended, the provisions of and the regulations of the Commission, and has the Act, and the regulations of the Commission, and been operating under a provisional license since has been operating under a provisional license since December 22, 1969; December 22, 1969; Construction of the Dresden Nuclear Power Station, Unit 2 (the facility) has been completed in conformity with Construction Permit No.

CPPR-18 and the application, as amended, the provisions of the Act, and the regulations of the Commission, and has been operating under a provisional license since December 22, 1969; Basis This license finding is proposed for deletion in its entirety. Decommissioning of Dresden, Unit 2, is not dependent on the regulations that governed construction of the facility.

Finding 1.c Current Proposed Actions have been identified and have been or will Actions have been identified and have been or will be taken with respect to (1) managing the effects of be taken with respect to (1) managing the effects aging during the period of extended operation on of aging during the period of extended operation the functionality of structures and components that on the functionality of structures and components have been identified to require review under that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review analyses that have been identified to require under 10 CFR 54.21(c), such that there is review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized reasonable assurance that the activities authorized by this renewed operating license will continue to by this renewed operating license will continue to be conducted in accordance with the current be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for licensing basis, as defined in 10 CFR 54.3, for Dresden Nuclear Power Station, Unit 2 (facility or Dresden Nuclear Power Station, Unit 2 (facility or plant), and that any changes made to the plant's plant), and that any changes made to the plant's current licensing basis in order to comply with current licensing basis in order to comply with 10 CFR 54.29(a) are in accord with the Act and the 10 CFR 54.29(a) are in accord with the Act and the Commission's regulations; Commission's regulations; Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Page 7 of 85

Attachment 1 Evaluation of Proposed Changes Finding 1.D Current Proposed The facility will operate in conformity with the The facility will operate be maintained in application, as amended, the provisions of the Act, conformity with the application, as amended, the and the regulations of the Commission (except as provisions of the Act, and the regulations of the exempted from compliance in Section 2.D below); Commission (except as exempted from compliance in Section 2.D below);

Basis This license finding is proposed for revision to reflect a more accurate description of the future requirements. Reference to exemptions in Section 2.D is proposed for deletion because Section 2.D is proposed for elimination in its entirety. Since the Dresden, Unit 2, license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel as provided in 10 CFR 50.82(a)(2), the removal of the operating description provides accuracy in the 10 CFR 50 license description. Therefore, the change is consistent with the requirements associated with the decommissioning facility.

Finding 1.E Current Proposed There is reasonable assurance (i) that the activities There is reasonable assurance (i) that the activities authorized by this renewed operating license can authorized by this renewed operating license can be conducted without endangering the health and be conducted without endangering the health and safety of the public and (ii) that such activities will safety of the public and (ii) that such activities will be conducted in compliance with the Commission's be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as regulations set forth in 10 CFR Chapter I (except exempted from compliance in Section 2.D. below); as exempted from compliance in Section 2.D.

below);

Basis This license finding is proposed for revision to reflect the change from an operating license to a non-operating license. After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Reference to exemptions in Section 2.D is proposed for deletion because Section 2.D is proposed for elimination in its entirety.

Finding 1.F Current Proposed Exelon Generation Company, LLC, is technically Exelon Generation Company, LLC, is technically and and financially qualified to engage in the activities financially qualified to engage in the activities authorized by this renewed operating license in authorized by this renewed operating license in accordance with the rules and regulations of the accordance with the rules and regulations of the Commission; Commission; Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Page 8 of 85

Attachment 1 Evaluation of Proposed Changes Finding 1.H Current Proposed The issuance of this renewed operating license will The issuance of this renewed operating license will not be inimical to the common defense and security not be inimical to the common defense and security or to the health and safety of the public; or to the health and safety of the public; Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding 1.I Current Proposed After weighing the environmental, economic, After weighing the environmental, economic, technical, and other benefits of the facility against technical, and other benefits of the facility against environmental and other costs and considering environmental and other costs and considering available alternatives, the issuance of this Renewed available alternatives, the issuance of this Renewed Facility Operating License No. DPR-19 is in Operating License No. DPR-19 is in accordance with accordance with 10 CFR Part 51 of the 10 CFR Part 51 of the Commission's regulations and Commission's regulations and all applicable all applicable requirements have been satisfied.; and requirements have been satisfied; and Basis This license finding is being revised to eliminate the reference to "Operating," and to end the list of findings as Finding 1.J is proposed for deletion in its entirety. After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding 1.J Current Proposed The receipt, possession, and use of source, Deleted. The receipt, possession, and use of source, byproduct and special nuclear material as byproduct and special nuclear material as authorized authorized by this renewed operating license will be by this renewed operating license will be in in accordance with the Commission's regulations in accordance with the Commission's regulations in 10 10 CFR Parts 30, 40 and 70. CFR Parts 30, 40 and 70.

Basis This license finding is proposed for deletion in its entirety. The Commission's finding regarding receipt, possession, and use of byproduct, source, and special nuclear material is not dependent on decommissioning of the facility. Additionally, possession and use of byproduct, source, and special nuclear material at Dresden during decommissioning activities is covered by License Condition 2.B, which will remain in effect. Therefore, License Finding 1.J is not needed General Paragraph 2 Current Proposed On the basis of the foregoing findings regarding this On the basis of the foregoing findings regarding this facility, Facility Operating License No. DPR-19, facility, Facility Operating License No. DPR-19, issued February 20, 1991, is superseded by issued February 20, 1991, is superseded by Renewed Facility Operating License No. DPR-19, Renewed Facility Operating License No. DPR-19, Page 9 of 85

Attachment 1 Evaluation of Proposed Changes which is hereby issued to Exelon Generation which is hereby issued to Exelon Generation Company, LLC, to read as Follows: Company, LLC, to read as Follows:

Basis This general paragraph is being revised to eliminate the reference to "Operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

General Paragraph 2.A Current Proposed This renewed operating license applies to the This renewed operating license applies to the Dresden Nuclear Power Station, Unit 2, a boiling Dresden Nuclear Power Station, Unit 2, a water reactor and associated equipment (the permanently defueled boiling water reactor and facility). The facility is located in Grundy County, associated equipment (the facility). The facility is Illinois, and is described in the licensee's Updated located in Grundy County, Illinois, and is described in Final Safety Analysis Report, as supplemented and the licensee's Updated Final Safety Analysis Report, amended, and in the licensee's Environmental as supplemented and amended, and in the Report, as supplemented and amended. licensee's Environmental Report, as supplemented and amended.

Basis This general paragraph is being revised to eliminate the reference to "Operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Additionally, descriptive language related to the permanently defueled condition of the Unit 2 reactor is proposed to be added.

Finding 2.B.(1)

Current Proposed Exelon Generation Company, LLC, pursuant to Exelon Generation Company, LLC, pursuant to Section 104b of the Act and 10 CFR Part 50, Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities", "Licensing of Production and Utilization Facilities", to to possess, use, and operate the facility at the possess, use, and operate and use the facility as designated location in Grundy, County, Illinois, in required for irradiated fuel storage at the accordance with the procedures and limitations set designated location in Grundy, County, Illinois, in forth in this renewed operating license; accordance with the procedures and limitations set forth in this renewed operating license; Basis This license finding is proposed for revision to reflect the change from an operating license to one that prohibits operating the reactor pursuant to 10 CFR 50.82(a)(2). As such, the facility would remain authorized to possess the existing irradiated fuel and use the facility as required to support safe storage of the irradiated fuel (e.g., the SFP) during the decommissioning period, in accordance with the specified limitations for storage.

Finding 2.B.(2)

Current Proposed Exelon Generation Company, LLC, pursuant to the Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor use at any time special nuclear materials that were fuel, in accordance with the limitations for storage used as reactor fuel, in accordance with the Page 10 of 85

Attachment 1 Evaluation of Proposed Changes and amounts required for reactor operation, as limitations for storage and amounts required for described in the Updated Final Safety Analysis reactor operation, as described in the Updated Final Report, as supplemented and amended; Safety Analysis Report, as supplemented and amended; Basis The proposed change to this license finding removes the authorization for receipt and use of special nuclear materials (SNM) as reactor fuel. It eliminates the reference to use of the SNM for reactor operations and limits the possession of SNM, to SNM "that were used" as reactor fuel at Dresden.

Pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 license for Dresden, Unit 2, will no longer authorize operation of the reactor. As such, Exelon has no need to receive SNM in the form of reactor fuel and cannot use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM "that were used" as reactor fuel is necessary as Exelon currently possesses the irradiated reactor fuel that was used for the past operations of the Dresden, Unit 2, reactor.

Finding 2.B.(3)

Current Proposed Exelon Generation Company, LLC, pursuant to the Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source possess and use at any time any byproduct, source and special nuclear material as sealed neutron and special nuclear material as sealed neutron sources for reactor startup, sealed sources for sources for reactor startup, or sealed sources for reactor instrumentation and radiation monitoring reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in equipment calibration, and as fission detectors in amounts as required; amounts as required to possess any byproduct, source and special nuclear material such as sealed neutron sources previously used for reactor startup or reactor instrumentation, and fission detectors; Basis The requirements regarding receipt of sealed neutron sources for reactor startup and nuclear instrumentation is proposed for deletion from this license finding. This license finding is revised to reflect authorization only for continued possession of those sources previously used for reactor startups and fission detection, produced as a byproduct, and the continued receipt and possession of those required for radiation monitoring equipment calibration. After the certifications required by 10 CFR 50.82(a)(1) have been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the use of startup sources and fission detectors will no longer be needed. The changes are consistent with the requirements associated with a permanently shutdown and defueled condition. The use of sources for radiation monitoring equipment calibration will continue to be required.

Finding 2.B.(5)

Current Proposed Exelon Generation Company, LLC, pursuant to the Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special but not separate, such byproduct and special nuclear nuclear materials as may be produced by the materials as may be that were produced by the operation of the facility. operation of the facility.

Basis This license finding is proposed for revision to allow possession of byproduct and SNM that were produced during operation of the reactor, but not allow the separation of material. After the certification required by Page 11 of 85

Attachment 1 Evaluation of Proposed Changes 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2). Therefore, the changes are consistent with the requirements associated with a permanently shutdown and defueled condition.

General Paragraph 2.C Current Proposed This renewed operating license shall be deemed to This renewed operating license shall be deemed to contain and is subject to the conditions specified in contain and is subject to the conditions specified in the following Commission regulations set forth in the following Commission regulations set forth in 10 CFR Chapter I; is subject to all applicable 10 CFR Chapter I; is subject to all applicable provisions of the Act and to the rules, regulations, provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in and orders of the Commission now or hereafter in effect; and is subject to the additional conditions effect; and is subject to the additional conditions specified or incorporated below: specified or incorporated below:

Basis This paragraph is proposed for revision to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2.C.(1) - Maximum Power Level Current Proposed Deleted. The licensee is authorized to operate the The licensee is authorized to operate the facility at facility at steady state reactor core power levels not steady state reactor core power levels not in excess in excess of 2957 megawatts thermal (100 percent of 2957 megawatts thermal (100 percent rated rated power) in accordance with the conditions power) in accordance with the conditions specified specified herein.

herein.

Basis This license condition is proposed for deletion in its entirety. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel has been submitted to the NRC pursuant to 10 CFR 50.82(a)(1)(i) and (ii) for Dresden, Unit 2, NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the reactor or emplacement of fuel into the reactor vessel under the 10 CFR 50 license. With the submittal of the certification in accordance with 10 CFR 50.82(a)(1), Exelon will no longer be authorized to operate the facility; therefore, this license condition is not needed.

License Condition 2.C.(2) - Technical Specifications Current Proposed The Technical Specifications contained in The Permanently Defueled Technical Specifications Appendix A, as revised through Amendment contained in Appendix A, as revised through No. 270, are hereby incorporated into this renewed Amendment No. [XXX], are hereby incorporated into operating license. The licensee shall operate the this renewed operating license. The licensee shall facility in accordance with the Technical operate maintain the facility in accordance with the Specifications. Permanently Defueled Technical Specifications.

Basis This license condition is revised to reflect the permanently shutdown and defueled condition of the facility.

Also changed is the designation from the licensee operating, to maintaining the facility. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessels Page 12 of 85

Attachment 1 Evaluation of Proposed Changes are submitted to the NRC pursuant to 10 CFR 50.82(a)(1)(i) and (ii) for Dresden, Unit 2, NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the reactor or emplacement of fuel into the reactor vessel under the 10 CFR 50 license.

License Condition 2.C.(3)

Current Proposed Operation in the coastdown mode is permitted to Deleted.Operation in the coastdown mode is 40% power. permitted to 40% power.

Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 2.C.(4)

Current Proposed The valves in the equalizer piping between the Deleted.The valves in the equalizer piping between recirculation loop shall be closed at all times during the recirculation loop shall be closed at all times reactor operation. during reactor operation.

Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been removed from the reactor, 10 CFR 50.82(a)(2) prohibits operation of the Dresden reactor since the certifications described therein will have been submitted. With operation of the reactor prohibited, the recirculation loops will no longer be required and as such, the equalizer valve restriction imposed by License Condition 2.C.(4) is no longer needed.

License Condition 2.C.(5)

Current Proposed The licensee shall maintain the commitments made Deleted. The licensee shall maintain the in response to the March 14, 1983, NUREG-0737 commitments made in response to the March 14, Order, subject to the following provision: 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the The licensee may make changes to March 14, 1983, NUREG-0737 Order commitments made in response to the March without prior approval of the Commission 14, 1983, NUREG-0737 Order without prior as long as the change would be permitted approval of the Commission as long as the without NRC approval, pursuant to the change would be permitted without NRC requirements of 10 CFR 50.59. Consistent approval, pursuant to the requirements of with this regulation, if the change results in 10 CFR 50.59. Consistent with this regulation, an Unreviewed Safety Question, a license if the change results in an Unreviewed Safety amendment shall be submitted to the NRC Question, a license amendment shall be staff for review and approval prior to submitted to the NRC staff for review and implementation of the change. approval prior to implementation of the change.

Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been Page 13 of 85

Attachment 1 Evaluation of Proposed Changes permanently removed from the reactor, reference to commitments related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 2.C.(6) - Surveillance Requirements Current Proposed The Surveillance Requirements contained in Deleted.The Surveillance Requirements Appendix A Technical Specifications and listed contained in Appendix A Technical Specifications below are not required to be performed immediately and listed below are not required to be performed upon implementation of Amendment No. 150: immediately upon implementation of Amendment No. 150:

a. Surveillance Requirement 4.1.A.2 - RPS a. Surveillance Requirement 4.1.A.2 - RPS Logic System Functional Test Logic System Functional Test

b. Surveillance Requirement 4.2.A.2 - Primary b. Surveillance Requirement

& Secondary Containment Logic System 4.2.A.2 - Primary & Secondary Functional Test Containment Logic System Functional Test

c. Surveillance Requirement c. Surveillance Requirement 4.2.J.2 - Feedwater Pump Trip Logic 4.2.J.2 - Feedwater Pump Trip Logic System Functional Test System Functional Test

d. Surveillance Requirement 4.6.F.1.b - Relief d. Surveillance Requirement Valve Logic System Functional Test 4.6.F.1.b - Relief Valve Logic System Functional Test

e. Surveillance Requirement e. Surveillance Requirement 4.9.A.9 - Simultaneous Diesel Generator 4.9.A.9 - Simultaneous Diesel Generator Start Start

f. Surveillance Requirement 4.9.A.10 - Diesel f. Surveillance Requirement Storage Tank Cleaning (Unit 3 and Unit 2/3 4.9.A.10 - Diesel Storage Tank only) Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the MODE 2 on the first plant startup following the fifteenth refueling outage (D2R15). fifteenth refueling outage (D2R15).

Basis This license condition is related to activities that have already occurred and is proposed for deletion in its entirety. Additionally, once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

Page 14 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 2.C.(7) - Additional Conditions Current Proposed The Additional Conditions contained in Appendix B, Deleted. The Additional Conditions contained in as revised through Amendment No. 191, are Appendix B, as revised through Amendment No.

hereby incorporated into this renewed operating 191, are hereby incorporated into this renewed license. The licensee shall operate the facility in operating license. The licensee shall operate the accordance with the Additional Conditions. facility in accordance with the Additional Conditions.

Basis This license condition is proposed for deletion in its entirety as the additional conditions in Appendix B to which it refers are proposed for deletion; therefore, retention of this license condition is unnecessary.

License Condition 2.C.(14)

Current Proposed Exelon Generation Company, LLC shall relocate Deleted. Exelon Generation Company, LLC shall certain Technical Specification requirements to EGC relocate certain Technical Specification requirements controlled documents upon implementation of the to EGC controlled documents upon implementation Amendment No. 185. The items and appropriate of the Amendment No. 185. The items and documents are as described in Table LA, "Removal appropriate documents are as described in Table LA, of Details Matrix," and Table R, "Relocated "Removal of Details Matrix," and Table R, "Relocated Specifications," that are attached to the NRCs Specifications," that are attached to the NRCs Safety Evaluation enclosed with Amendment Safety Evaluation enclosed with Amendment No. 185. No. 185.

Basis This license condition resulted from the implementation of Improved TS for Dresden, Unit 2, (i.e., implementation of NUREG-1433) and is proposed for deletion in its entirety. It was related to the relocation of items from the TS to the Technical Requirements Manual, and is related to activities that have already occurred. After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2); therefore, retention of these requirements are no longer necessary for the maintenance of the facility.

License Condition 2.C.(15)

Current Proposed The schedule for performing Surveillance Deleted. The schedule for performing Surveillance Requirements (SRs) that are new or revised in Requirements (SRs) that are new or revised in Amendment No. 185 shall be as follows: Amendment No. 185 shall be as follows:

For SRs that are new in this amendment, For SRs that are new in this amendment, the first performance is due at the end of the first performance is due at the end of the first surveillance interval that begins on the first surveillance interval that begins on the date of implementation of Amendment the date of implementation of Amendment No. 185. No. 185.

For SRs that existed prior to this For SRs that existed prior to this amendment whose intervals of amendment whose intervals of performance are being reduced, the first performance are being reduced, the first reduced surveillance interval begins upon reduced surveillance interval begins upon Page 15 of 85

Attachment 1 Evaluation of Proposed Changes completion of the first surveillance completion of the first surveillance performed after implementation of performed after implementation of Amendment No. 185. Amendment No. 185.

For SRs that existed prior to this For SRs that existed prior to this amendment that have modified acceptance amendment that have modified acceptance criteria, the first performance is due at the criteria, the first performance is due at the end of the first surveillance interval that end of the first surveillance interval that began on the date the surveillance was last began on the date the surveillance was last performed prior to the implementation of performed prior to the implementation of Amendment No. 185. Amendment No. 185.

For SRs that existed prior to this For SRs that existed prior to this amendment amendment whose intervals of whose intervals of performance are being performance are being extended, the first extended, the first extended surveillance extended surveillance interval begins upon interval begins upon completion of the last completion of the last surveillance surveillance performed prior to performed prior to implementation of implementation of Amendment No. 185.

Amendment No. 185.

Basis This license condition is related to activities that have already occurred and is proposed for deletion in its entirety. Additionally, once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 2.C.(16)

Current Proposed Following implementation of Amendment No. 185, Deleted. Following implementation of Amendment the reactor protection system trip setpoint for main No. 185, the reactor protection system trip setpoint steam isolation valve closure shall be maintained at for main steam isolation valve closure shall be the previous setpoint (less than or equal to 10% maintained at the previous setpoint (less than or closed) until startup after the first outage of equal to 10% closed) until startup after the first sufficient duration to change the setpoint. outage of sufficient duration to change the setpoint.

Basis This license condition is related to activities that have already occurred and is proposed for deletion in its entirety. Additionally, once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

Page 16 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 2.C.(17)

Current Proposed The license is amended to authorize changing the Deleted. The license is amended to authorize UFSAR to allow credit for containment changing the UFSAR to allow credit for containment overpressure as detailed below, to assure adequate overpressure as detailed below, to assure adequate Net Positive Suction Head is available for low Net Positive Suction Head is available for low pressure Emergency Core Cooling System pumps pressure Emergency Core Cooling System pumps following a design basis accident. following a design basis accident.

From (sec) To (sec) Credit (psig) From (sec) To (sec) Credit (psig)

Accident start 290 9.5 Accident start 290 9.5 290 5,000 4.8 290 5,000 4.8 5,000 30,000 6.6 5,000 30,000 6.6 30,000 40,000 6.0 30,000 40,000 6.0 40,000 45,500 5.4 40,000 45,500 5.4 45,500 52,500 4.9 45,500 52,500 4.9 52,500 60,500 4.4 52,500 60,500 4.4 60,500 70,000 3.8 60,500 70,000 3.8 70,000 84,000 3.2 70,000 84,000 3.2 84,000 104,000 2.5 84,000 104,000 2.5 104,000 136,000 1.8 104,000 136,000 1.8 136,000 Accident end 1.1 136,000 Accident end 1.1 Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been removed from the reactor, 10 CFR 50.82(a)(2) prohibits operation of the Dresden, Unit 2, reactor. With operation of the reactor prohibited, the Emergency Core Cooling System pumps will no longer be required for mitigation of design basis accidents and as such, the crediting of containment overpressure allowed by License Condition 2.C.(17) is no longer needed.

License Condition 2.C.(20)

Current Proposed Upon implementation of Amendment No. 226 Deleted. Upon implementation of Amendment No.

adopting TSTF-448, Revision 3, the determination 226 adopting TSTF-448, Revision 3, the of control room envelope (CRE) unfiltered air determination of control room envelope (CRE) inleakage as required by SR 3.7.4.4, in accordance unfiltered air inleakage as required by SR 3.7.4.4, with TS 5.5.14.c.(i), the assessment of CRE in accordance with TS 5.5.14.c.(i), the assessment habitability as required by Specification 5.5.14.c.(ii), of CRE habitability as required by Specification and the measurement of CRE pressure as required 5.5.14.c.(ii), and the measurement of CRE pressure by Specification 5.5.14.d, shall be considered met. as required by Specification 5.5.14.d, shall be Following implementation: considered met. Following implementation:

(a) The first performance of SR 3.7.4.4, in (a) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), accordance with Specification 5.5.14.c.(i),

shall be within the specified Frequency of shall be within the specified Frequency of 6 years, plus the 18-month allowance of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from January 1997, SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer the date of the most recent successful tracer gas test, as stated in the December 9, 2003 gas test, as stated in the December 9, 2003 Page 17 of 85

Attachment 1 Evaluation of Proposed Changes letter response to Generic Letter 2003-01, or letter response to Generic Letter 2003-01, or within the next 18 months if the time period within the next 18 months if the time period since the most recent successful tracer gas since the most recent successful tracer gas test is greater than 6 years. test is greater than 6 years.

(b) The first performance of the periodic (b) The first performance of the periodic assessment of CRE habitability, Specification assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured 9-month allowance of SR 3.0.2, as measured from January 1997, the date of the most from January 1997, the date of the most recent successful tracer gas test, as stated in recent successful tracer gas test, as stated in the December 9, 2003 letter response to the December 9, 2003 letter response to Generic Letter 2003-01, or within the next Generic Letter 2003-01, or within the next 9 months if the time period since the most 9 months if the time period since the most recent successful tracer gas test is greater recent successful tracer gas test is greater than 3 years. than 3 years.

(c) The first performance of the periodic (c) The first performance of the periodic measurement of CRE pressure, measurement of CRE pressure, Specification Specification 5.5.14.d, shall be within 5.5.14.d, shall be within 24 months, plus the 6 24 months, plus the 6 months allowed by months allowed by SR 3.0.2, as measured SR 3.0.2, as measured from the date of the from the date of the most recent successful most recent successful pressure pressure measurement test, or within 6 measurement test, or within 6 months if not months if not performed previously.

performed previously.

Basis This license condition is proposed for deletion in its entirety. The proposed change removes the requirements of TSTF-448 that involve assessing the Control Room Envelope (CRE) Habitability at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. These assessments were completed in accordance with the schedule specified in the license condition.

Exelon performed an analysis for an FHA in the SFP for dose results for the control room (CR) after permanent shutdown. The calculation accounts for radioactive material inventory in the most recently irradiated elements in the SFP after 48 days of decay following permanent shutdown. For the analysis, Exelon took no credit for filtered recirculation of the control room air. The results of the calculation showed that the dose consequences to occupants in the control room were below the 10 CFR 50.67 dose limit of 5 rem. Based on the fact that the dose at the CR is less than the 10 CFR 50.67 dose limit and that no credit was taken for CR filtered recirculation, the CRE Habitability Program is not required to provide airborne radiological protection for the control room operators. This submittal also proposes to remove Technical Specification 3.7.4 for the control room emergency ventilation system and Technical Specification 5.5.14 for the CRE Habitability Program. Since TS 3.7.4 and TS 5.5.14 are no longer necessary, this license condition is no longer needed; therefore, it is proposed for deletion.

Proposed License Condition 2.C.(22)

Current Proposed Handling of irradiated fuel in the spent fuel pool

[None] will not be permitted following implementation of the PDTS until a minimum of 48 days following permanent shutdown.

Page 18 of 85

Attachment 1 Evaluation of Proposed Changes Basis Exelon is proposing a new license condition such that initial system abandonment activities may be started expeditiously after the permanent removal of fuel from the reactor vessel. By applying this new license condition, Exelon will be able to remove the TS requirements associated with those systems that perform mitigative actions assumed in the FHA analyses by precluding the possibility of an FHA until after the 48-day decay period assumed in the post permanent shutdown FHA has elapsed.

Once the reactor has been permanently defueled with all irradiated fuel placed in the SFP and the certifications submitted in accordance with 10 CFR 50.82, power operation or emplacement of fuel in the reactor will not be allowed. Therefore, all DBAs associated with power operations or fuel handling in the reactor will no longer be applicable, which provides the basis for removal of the Safety Limits and most of the Limiting Conditions for Operation.

In order to implement the PDTS prior to the 48-day decay time assumed in the post permanent shutdown FHA analysis, Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 days after permanent shutdown through the imposition of the proposed license condition. This effectively prevents an FHA from occurring until after the 48-day decay period has elapsed.

License Condition 2.D Current Proposed The facility has been granted certain exemptions Deleted. The facility has been granted certain from the requirements of Section III.G of exemptions from the requirements of Section III.G Appendix R to 10 CFR Part 50, "Fire Protection of Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems fire protection features for ensuring the systems and associated circuits used to achieve and and associated circuits used to achieve and maintain safe shutdown are free of fire damage. maintain safe shutdown are free of fire damage.

These exemptions were granted and sent to the These exemptions were granted and sent to the licensee in letters dated February 2, 1983, licensee in letters dated February 2, 1983, September 28, 1987, July 6, 1989, and September 28, 1987, July 6, 1989, and August 15, 1989. August 15, 1989.

In addition, the facility has been granted certain In addition, the facility has been granted certain exemptions from Sections II and III of Appendix J to exemptions from Sections II and III of Appendix J to 10 CFR Part 50, "Primary Reactor Containment 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Leakage Testing for Water-Cooled Power Reactors." This section contains leakage test Reactors." This section contains leakage test requirements, schedules and acceptance criteria for requirements, schedules and acceptance criteria for tests of the leak-tight integrity of the primary reactor tests of the leak-tight integrity of the primary reactor containment and systems and components which containment and systems and components which penetrate the containment. These exemptions penetrate the containment. These exemptions were granted and sent to the licensee in a letter were granted and sent to the licensee in a letter dated June 25, 1982. dated June 25, 1982.

These exemptions granted pursuant to These exemptions granted pursuant to 10 CFR 50.12 are authorized by law, will not 10 CFR 50.12 are authorized by law, will not present present an undue risk to the public health and an undue risk to the public health and safety, and are safety, and are consistent with the common consistent with the common defense and security.

Page 19 of 85

Attachment 1 Evaluation of Proposed Changes defense and security. With these exemptions, the With these exemptions, the facility will operate, to the facility will operate, to the extent authorized herein, extent authorized herein, in conformity with the in conformity with the application, as amended, the application, as amended, the provisions of the Act, provisions of the Act, and the rules and regulations and the rules and regulations of the Commission.

of the Commission.

Basis This license condition is proposed for deletion in its entirety. This license condition documents specific exemptions from 10 CFR 50 as approved by the NRC. Specifically, exemption from the requirements of 10 CFR 50, Appendix R, Section III.G and 10 CFR 50, Appendix J.

The requirements of 10 CFR 50, Appendix R required to mitigate the consequences of design basis accidents under post-fire conditions, limit fire damage to systems required to achieve, and maintain safe shutdown conditions. Once Dresden, Unit 2 has permanently ceased operation and certified that fuel has been permanently removed from the reactor, the requirements of Appendix R will no longer apply.

Therefore, Exelon will no longer need these exemptions to Appendix R. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. The regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled facility is not needed. The fire protection program will be revised to take into account the facility conditions and activities during decommissioning. Exelon will continue to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment as a result of activities at the Dresden facility.

Similarly, following the submittal of the certification under 10 CFR 50.82(a)(1) has been submitted to the NRC for Dresden, Unit 2, the provisions of 10 CFR 50, Appendix J, will no longer apply, since, as stated in 10 CFR 50.54(o), "Condition of Licenses," "Primary reactor containments for water cooled power reactors, other than facilities for which the certifications required under §§ 50.82(a)(1) or 52.110(a)(1) of this chapter have been submitted, shall be subject to the requirements set forth in Appendix J to this part."

License Condition 2.E Current Proposed The licensee shall implement and maintain in effect Deleted. The licensee shall implement and all provisions of the approved fire protection maintain in effect all provisions of the approved fire program as described in the Updated Final Safety protection program as described in the Updated Analysis Report for the facility and as approved in Final Safety Analysis Report for the facility and as the Safety Evaluation Reports dated March 22, approved in the Safety Evaluation Reports dated 1978 with supplements dated December 2, 1980, March 22, 1978 with supplements dated December and February 12, 1981; January 19, 1983; July 17, 2, 1980, and February 12, 1981; January 19, 1983; 1987; September 28, 1987; and January 5, 1989, July 17, 1987; September 28, 1987; and January 5, subject to the following provision: 1989, subject to the following provision:

The licensee may make changes to the The licensee may make changes to the approved fire protection program without approved fire protection program without prior approval of the Commission only if prior approval of the Commission only if those changes would not adversely affect those changes would not adversely affect the ability to achieve and maintain safe the ability to achieve and maintain safe shutdown in the event of a fire. shutdown in the event of a fire.

Basis This license condition is proposed for deletion in its entirety consistent with the restriction of 10 CFR 50.82(a)(2) that Dresden, Unit 2, is no longer authorized to operate or emplace fuel in the reactor.

Page 20 of 85

Attachment 1 Evaluation of Proposed Changes This license condition for making changes to the Fire Protection Program is no longer required to assure fire safety by maintaining the ability to achieve and maintain safe shutdown in the event of a fire.

License Condition 2.E, which is based on maintaining an operational fire protection program, in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shutdown of the reactors in the event of a fire, is no longer applicable for Dresden, Unit 2. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during plant decommissioning. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled plant facility is not required.

License Condition 2.I - Updated Final Safety Analysis Report Current Proposed The Exelon Generation Company, LLC Updated Deleted. The Exelon Generation Company, LLC Final Safety Analysis Report supplement, submitted Updated Final Safety Analysis Report supplement, pursuant to 10 CFR 54.21(d), describes certain submitted pursuant to 10 CFR 54.21(d), describes future activities to be completed prior to the period certain future activities to be completed prior to the of extended operation. The Exelon Generation period of extended operation. The Exelon Company, LLC shall complete these activities no Generation Company, LLC shall complete these later than December 22, 2009, and shall notify the activities no later than December 22, 2009, and NRC in writing when implementation of these shall notify the NRC in writing when implementation activities is complete and can be verified by NRC of these activities is complete and can be verified inspection. by NRC inspection.

The Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement, as revised, shall be included in the supplement, as revised, shall be included in the next next scheduled update to the Updated Final Safety scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4) Analysis Report required by 10 CFR 50.71 (e)(4) following issuance of this renewed license. Until following issuance of this renewed license. Until that that update is complete, Exelon Generation update is complete, Exelon Generation Company, Company, LLC may make changes to the programs LLC may make changes to the programs and and activities described in the supplement without activities described in the supplement without prior prior Commission approval, provided that Exelon Commission approval, provided that Exelon Generation Company, LLC evaluates such change Generation Company, LLC evaluates such change pursuant to the criteria set forth in 10 CFR 50.59 pursuant to the criteria set forth in 10 CFR 50.59 and and otherwise complies with the requirements in otherwise complies with the requirements in that that section. section.

Basis This license condition imposed program requirements necessary for the Dresden, Unit 2, to continue operation beyond December 22, 2009. After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Thus, the program requirements to support extended operation are not necessary.

License Condition 2.J Current Proposed All capsules in the reactor vessel that are removed Deleted. All capsules in the reactor vessel that are and tested must meet the test procedures and removed and tested must meet the test procedures reporting requirements of ASTM E 185 82 to the and reporting requirements of ASTM E 185 82 to the Page 21 of 85

Attachment 1 Evaluation of Proposed Changes extent practicable for the configuration of the extent practicable for the configuration of the specimens in the capsule. Any changes to the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to capsules, must be approved by the NRC prior to implementation. All capsules placed in storage implementation. All capsules placed in storage must must be maintained for future insertion. be maintained for future insertion.

Basis This license condition was issued concurrent with the Renewed Facility Operating License on October 28, 2004. This license condition is described in Section 1.7, "Summary of Proposed License Conditions," of NUREG-1796, "Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2" issued October 2004 (Reference 6).

This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Continued implementation of the applicable surveillance capsule testing and reporting requirements are no longer necessary for Dresden, Unit 2 because: (a) Exelon has decided to cease power operations of Dresden, Unit 2 and (b) from a fracture toughness perspective, the Dresden, Unit 2, RPV will cease to be exposed to further irradiation by high energy neutrons or subjected to any high thermal stress environments, as induced by operating the RCS at an elevated temperature. Further, 10 CFR 50.60(a) stipulates that reactor facilities for which the certifications required under § 50.82(a)(1) have been submitted, no longer need to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in Appendices G and H.

The physical and radiological control of the remaining surveillance capsules that are located in the RPV will be managed in accordance with the applicable radiological control requirements of 10 CFR 20 and with any applicable security or physical protection requirements for components in either 10 CFR 37 or 10 CFR 73.

Therefore, the removal, testing, reporting, and storage requirements for reactor vessel surveillance capsules and their test specimens do not need to be implemented further once Dresden, Unit 2, permanently ceases power operations. There will no longer be any need to remove the remaining surveillance capsules from the RPV or perform material testing of the test specimens in those capsules. As such, deletion of this license condition is appropriate. Any corresponding commitments in the Dresden, Units 2 and 3, UFSAR will also be deleted under the provisions of 10 CFR 50.59 upon NRC approval of this proposed license amendment request.

License Condition 3 Current Proposed This renewed operating license is effective as of the This renewed operating license is effective as of the date of issuance and shall expire at midnight on date of issuance and shall expire at midnight on December 22, 2029. December 22, 2029 is effective until the Commission notifies the licensee in writing that the license is terminated.

Basis This License Condition is revised to reflect the permanently defueled condition of the facility. Once Dresden has permanently ceased operation and Exelon has certified that fuel has been removed from the reactors, 10 CFR 50.82(a)(2) prohibits operation of the Dresden reactors. Reference to an operating state for the facility is deleted and this license condition is proposed for further revision in accordance with 10 CFR 50.51(b) in that the license authorizes ownership and possession by Exelon until the Commission notifies the licensee in writing that the license is terminated.

Page 22 of 85

Attachment 1 Evaluation of Proposed Changes Appendix B - Additional Conditions - Facility Operating License No. DPR-19 Current Proposed Deleted Basis This Appendix to the RFOL contains additional conditions that are related to activities that have already occurred and are proposed for deletion in their entirety. Once Dresden, Unit 2, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to these activities that are associated with the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

Detailed Description of the Proposed Change to the Dresden, Unit 3 RFOL RFOL Title Current Proposed Renewed Facility Operating License Renewed Facility Operating License Basis The License title is modified to eliminate the reference to "Operating." After certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding b Current Proposed Actions have been identified and have been or will Actions have been identified and have been or will be taken with respect to (1) managing the effects of be taken with respect to (1) managing the effects aging during the period of extended operation on of aging during the period of extended operation the functionality of structures and components that on the functionality of structures and components have been identified to require review under that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review analyses that have been identified to require Page 23 of 85

Attachment 1 Evaluation of Proposed Changes under 10 CFR 54.21(c), such that there is review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized reasonable assurance that the activities authorized by this renewed operating license will continue to by this renewed operating license will continue to be conducted in accordance with the current be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for licensing basis, as defined in 10 CFR 54.3, for Dresden Nuclear Power Station, Unit 3 (facility or Dresden Nuclear Power Station, Unit 3 (facility or plant), and that any changes made to the plant's plant), and that any changes made to the plant's current licensing basis in order to comply with current licensing basis in order to comply with 10 CFR 54.29(a) are in accord with the Act and the 10 CFR 54.29(a) are in accord with the Act and the Commission's regulations; Commission's regulations; Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding d Current Proposed The Dresden Nuclear Power Station Unit 3 Deleted; The Dresden Nuclear Power Station Unit 3 (the facility) has been substantially completed in (the facility) has been substantially completed in conformity with Provisional Construction Permit conformity with Provisional Construction Permit No. CPPR-22, the application, the provisions of the No. CPPR-22, the application, the provisions of the Act and the rules and regulations of the Commission; Act and the rules and regulations of the Commission; Basis This license finding is proposed for deletion in its entirety. Decommissioning of Dresden, Unit 3, is not dependent on the regulations that governed construction of the facility.

Finding e Current Proposed The facility will operate in conformity with the The facility will operate be maintained in application, as amended, the provisions of the Act, conformity with the application, as amended, the and the regulations of the Commission; provisions of the Act, and the regulations of the Commission; Basis This license finding is proposed for revision to reflect a more accurate description of the future requirements. Since the Dresden, Unit 3, license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel as provided in 10 CFR 50.82(a)(2),

the removal of the operating description provides accuracy in the 10 CFR 50 license description.

Therefore, the change is consistent with the requirements associated with the decommissioning plant.

Finding f Current Proposed There is reasonable assurance (i) that the facility There is reasonable assurance (i) that the facility can be operated at power levels not in excess of can be operated at power levels not in excess of 2957 megawatts (thermal) in accordance with this 2957 megawatts (thermal) maintained in license without endangering the health and safety accordance with this license without endangering of the public, and (ii) that such activities will be the health and safety of the public, and (ii) that Page 24 of 85

Attachment 1 Evaluation of Proposed Changes conducted in compliance with the rules and such activities will be conducted in compliance with regulations of the Commission; the rules and regulations of the Commission; Basis This license finding is being revised to eliminate the reference to operation of the facility, and modified to reflect that future activities will be related to the maintenance of the facility to similarly protect the health and safety of the public in accordance with the rules and regulations of the Commission. This is due to the fact that after the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding g Current Proposed Exelon Generation Company, LLC, is technically Exelon Generation Company, LLC, is technically and and financially qualified to engage in the activities financially qualified to engage in the activities authorized by this renewed operating license in authorized by this renewed operating license in accordance with the rules and regulations of the accordance with the rules and regulations of the Commission; Commission; Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Finding i Current Proposed The issuance of this renewed operating license will The issuance of this renewed operating license will not be inimical to the common defense and security not be inimical to the common defense and security or to the health and safety of the public; and or to the health and safety of the public; and Basis This license finding is being revised to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Page 25 of 85

Attachment 1 Evaluation of Proposed Changes Finding j Current Proposed After weighing the environmental, economic, After weighing the environmental, economic, technical, and other benefits of the facility against technical, and other benefits of the facility against environmental and other costs and considering environmental and other costs and considering available alternatives, the issuance of this Renewed available alternatives, the issuance of this Renewed Facility Operating License No. DPR-25 is in Operating License No. DPR-25 is in accordance with accordance with 10 CFR Part 51 of the 10 CFR Part 51 of the Commission's regulations and Commission's regulations and all applicable all applicable requirements have been satisfied; and requirements have been satisfied; and Basis This License Finding is being revised to eliminate the reference to "Operating," After certifications required by 10 CFR 50.82(a)(1) have been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition - General Paragraph Current Proposed On the basis of the foregoing findings regarding this On the basis of the foregoing findings regarding this facility, Facility Operating License No. DPR-25, facility, Facility Operating License No. DPR-25, issued January 12, 1971, is superseded by issued January 12, 1971, is superseded by Renewed Renewed Facility Operating License No. DPR-25, Facility Operating License No. DPR-25, which is which is hereby issued to Exelon Generation hereby issued to Exelon Generation Company, LLC Company, LLC (EGC or the licensee), to read as (EGC or the licensee), to read as Follows:

Follows:

Basis This general paragraph is being revised to eliminate the reference to "Operating." After certifications required by 10 CFR 50.82(a)(1) have been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

General Paragraph 1 Current Proposed This renewed operating license applies to the This renewed operating license applies to the Dresden Nuclear Power Station, Unit 3, a single Dresden Nuclear Power Station, Unit 3, a cycle, boiling, light water reactor and electric permanently defueled, single cycle, boiling, light generating equipment (the facility). The facility is water reactor and electric generating equipment (the located at the Dresden Nuclear Power Station in facility). The facility is located at the Dresden Grundy County, Illinois, and is described in the Nuclear Power Station in Grundy County, Illinois, licensee's "Safety Analysis Report," as and is described in the licensee's "Safety Analysis supplemented and amended (Amendment Nos. 8 Report," as supplemented and amended through 24). (Amendment Nos. 8 through 24).

Basis This general paragraph is being revised to eliminate the reference to "Operating." After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Additionally, descriptive language related to the permanently defueled condition of the Unit 3 reactor is proposed to be added to this license condition.

Page 26 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 2.A Current Proposed Pursuant to Section 104b of the Act and Pursuant to Section 104b of the Act and 10 CFR Part 10 CFR Part 50, "Licensing of Production and 50, "Licensing of Production and Utilization Utilization Facilities", to possess, use, and operate Facilities", to possess, use, and operate and use the the facility as a utilization facility at the designated facility as a utilization facility required for irradiated location at the Dresden Nuclear Power Station, in fuel storage at the designated location at the accordance with the procedures and limitations set Dresden Nuclear Power Station, in accordance with forth in this renewed operating license; the procedures and limitations set forth in this renewed operating license; Basis This license condition is proposed for revision to reflect the change from an operating license for a utilization facility, to one that is prohibited from operating the reactor pursuant to 10 CFR 50.82(a)(2). This includes the deletion of all occurrences of the word "operating." As such, the facility would remain authorized to possess the existing irradiated fuel and use the systems required to support safe fuel storage (e.g., the SFP) during the decommissioning period, in accordance with the specified limitations for irradiated fuel storage.

License Condition 2.B Current Proposed Pursuant to the Act and 10 CFR Part 70, to receive, Exelon Generation Company, LLC, pursuant to the possess and use at any time special nuclear Act and 10 CFR Part 70, to receive, possess and material, not including plutonium, as reactor fuel, in use at any time special nuclear material, not accordance with the limitations for storage and including plutonium that was used as reactor fuel, in amounts required for reactor operation, as accordance with the limitations for storage and described in the Final Safety Analysis Report, as amounts required for reactor operation, as described supplemented and amended as of in the Final Safety Analysis Report, as supplemented September 3, 1976; and amended as of September 3, 1976; Basis The proposed change to this license condition removes the authorization for receipt and use of special nuclear material (SNM) as reactor fuel. It eliminates the reference to use of the SNM for reactor operations and limits the possession of SNM to SNM "that was used" as reactor fuel at Dresden. Pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 license for Dresden, Unit 3, will no longer authorize operation of the reactor. As such, Exelon has no need to receive SNM in the form of reactor fuel and cannot no longer use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM "that was used" as reactor fuel is necessary as Exelon will continue to possess the reactor fuel that was used for the past operations of the Dresden, Unit 3, reactor.

License Condition 2.C Current Proposed Pursuant to the Act and 10 CFR Parts 30, 40 and Pursuant to the Act and 10 CFR Parts 30, 40 and 70, 70, to receive, possess and use at any time any to receive, possess and use at any time any byproduct, source and special nuclear material as byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sealed neutron sources for reactor startup, or sealed sources for reactor instrumentation and radiation sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission monitoring equipment calibration, and as fission detectors in amounts as required; detectors in amounts as required to possess any byproduct, source and special nuclear material Page 27 of 85

Attachment 1 Evaluation of Proposed Changes such as sealed neutron sources previously used for reactor startup or reactor instrumentation, and fission detectors; Basis The requirements regarding receipt of sealed neutron sources for reactor startup and nuclear instrumentation is proposed for deletion from this license condition. This license condition is revised to reflect authorization only for continued possession of those sources previously used for reactor startups and fission detection, produced as a byproduct, and the receipt and possession of those required for radiation monitoring equipment calibration. After certifications required by 10 CFR 50.82(a)(1) have been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the use of startup sources and fission detectors will no longer be needed. The changes are consistent with the requirements associated with a permanently shutdown and defueled condition. The use of sources for radiation monitoring equipment calibration will continue to be required.

License Condition 2.E Current Proposed Pursuant to the Act and 10 CFR Parts 30 and 70, to Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and possess, but not separate, such byproduct and special nuclear materials as may be produced by special nuclear materials as may be that were the operation of the Dresden Nuclear Power produced by the operation of the Dresden Nuclear Station, Unit Nos. 1, 2, and 3. Power Station, Unit Nos. 1, 2, and 3.

Basis This license condition is proposed for revision to allow possession of byproduct and SNM that were produced during operation of the reactor, but not allow the separation of material. After the certification required by 10 CFR 50.82(a)(1) has been submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2). Therefore, the changes are consistent with the requirements associated with a permanently shutdown and defueled condition.

License Condition 2.F - Surveillance Requirements Current Proposed The Surveillance Requirements contained in Deleted. The Surveillance Requirements Appendix A Technical Specifications and listed contained in Appendix A Technical Specifications below are not required to be performed immediately and listed below are not required to be performed upon implementation of Amendment No. 145: immediately upon implementation of Amendment No. 145:

a. Surveillance Requirement 4.1.A.2 - RPS a. Surveillance Requirement 4.1.A.2 - RPS Logic System Functional Test Logic System Functional Test

b. Surveillance Requirement 4.2.A.2 - Primary b. Surveillance Requirement

& Secondary Containment Logic System 4.2.A.2 - Primary & Secondary Functional Test Containment Logic System Functional Test

c. Surveillance Requirement c. Surveillance Requirement 4.2.J.2 - Feedwater Pump Trip Logic 4.2.J.2 - Feedwater Pump Trip Logic System Functional Test System Functional Test

d. Surveillance Requirement 4.6.F.1.b - Relief d. Surveillance Requirement Valve Logic System Functional Test 4.6.F.1.b - Relief Valve Logic System Functional Test Page 28 of 85

Attachment 1 Evaluation of Proposed Changes

e. Surveillance Requirement e. Surveillance Requirement 4.9.A.9 - Simultaneous Diesel Generator 4.9.A.9 - Simultaneous Diesel Generator Start Start

f. Surveillance Requirement 4.9.A.10 - Diesel f. Surveillance Requirement Storage Tank Cleaning (Unit 3 and Unit 2/3 4.9.A.10 - Diesel Storage Tank only) Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14). fourteenth refueling outage (D3R14).

Basis This license condition is related to activities that have already occurred and is proposed for deletion in its entirety. Additionally, once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 3 Current Proposed This renewed operating license shall be deemed to This renewed operating license shall be deemed to contain and is subject to the conditions specified in contain and is subject to the conditions specified in the following Commission regulations: the following Commission regulations: 10 CFR Part 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, 20, Section 30.34 of 10 CFR Part 30, Section 40.41 Section 40.41 of 10 CFR Part 40, Sections 50.54 of 10 CFR Part 40, Sections 50.54 and 50.59 of and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 50, and Section 70.32 of 10 CFR 10 CFR Part 70; is subject to all applicable Part 70; is subject to all applicable provisions of the provisions of the Act and to the rules, regulations, Act and to the rules, regulations, and orders of the and orders of the Commission now or hereafter in Commission now or hereafter in effect; and is subject effect; and is subject to the additional conditions to the additional conditions specified or incorporated specified or incorporated below: below:

Basis This license condition is proposed for revision to eliminate the reference to "operating." After the certification required by 10 CFR 50.82(a)(1) is submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 3.A - Maximum Power Level Current Proposed The licensee is authorized to operate the facility at Deleted. The licensee is authorized to operate the steady state power levels not in excess of 2957 facility at steady state power levels not in excess of megawatts (thermal), except that the licensee shall 2957 megawatts (thermal), except that the licensee not operate the facility at power levels in excess of shall not operate the facility at power levels in excess five (5) megawatts (thermal), until satisfactory of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the completion of modifications and final testing of the station output transformer, the auto- station output transformer, the auto-depressurization depressurization interlock, and the feedwater interlock, and the feedwater system, as described in system, as described in the licensees telegrams; the licensees telegrams; dated February 26, 1971, dated February 26, 1971, have been verified in have been verified in writing by the Commission.

writing by the Commission.

Basis Page 29 of 85

Attachment 1 Evaluation of Proposed Changes This license condition is proposed for deletion in its entirety. Once the certification for permanent cessation of operations and permanent removal of fuel from the reactor vessel has been submitted to the NRC pursuant to 10 CFR 50.82(a)(1)(i) and (ii) for Dresden, Unit 3, NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the reactor or emplacement of fuel into the reactor vessel under the 10 CFR 50 license. With the submittal of the certification in accordance with 10 CFR 50.82(a)(1), Exelon will no longer be authorized to operate the facility; therefore, this license condition is not needed.

License Condition 3.B - Technical Specifications Current Proposed The Technical Specifications contained in The Permanently Defueled Technical Specifications Appendix A, as revised through Amendment contained in Appendix A, as revised through No. 263, are hereby incorporated into this renewed Amendment No. [XXX], are hereby incorporated into operating license. The licensee shall operate the this renewed operating license. The licensee shall facility in accordance with the Technical operate maintain the facility in accordance with the Specifications. Permanently Defueled Technical Specifications.

Basis This license condition is revised to reflect the permanently shutdown and defueled condition of the facility.

Also changed is the designation from the licensee operating, to maintaining the facility. 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)(i) and (ii) for Dresden, Unit 2, NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the reactor or emplacement of fuel into the reactor vessel under the 10 CFR 50 license.

License Condition 3.C Current Proposed The licensee shall make certain reports in The licensee shall make certain reports in accordance with the requirements of the Technical accordance with the requirements of the Specifications. Permanently Defueled Technical Specifications.

Basis This license condition is proposed for modification to reflect that once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, facility reports shall be made in accordance with the Permanently Defueled Technical Specifications consistent with the provisions of 10 CFR 50.82(a)(2).

Page 30 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 3.D Current Proposed The licensee shall keep facility operating records in The licensee shall keep facility operating records in accordance with the requirements of the Technical accordance with the requirements of the Specifications. Permanently Defueled Technical Specifications.

Basis This license condition is proposed for modification to reflect that once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, facility records shall be maintained in accordance with the Permanently Defueled Technical Specifications consistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 3.E - Restrictions Current Proposed Operation in the coastdown mode is permitted to Deleted. Operation in the coastdown mode is 40% power. permitted to 40% power.

Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to operation of the reactor in the coastdown mode would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 3.F Current Proposed The licensee shall maintain the commitments made Deleted. The licensee shall maintain the in response to the March 14, 1983, NUREG-0737 commitments made in response to the March 14, Order, subject to the following provision: 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the The licensee may make changes to March 14, 1983, NUREG-0737 Order commitments made in response to the March 14, without prior approval of the Commission 1983, NUREG-0737 Order without prior approval as long as the change would be permitted of the Commission as long as the change would without NRC approval, pursuant to the be permitted without NRC approval, pursuant to requirements of 10 CFR 50.59. Consistent the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in with this regulation, if the change results in an an Unreviewed Safety Question, a license Unreviewed Safety Question, a license amendment shall be submitted to the NRC amendment shall be submitted to the NRC staff staff for review and approval prior to for review and approval prior to implementation implementation of the change. of the change.

Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to commitments related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

Page 31 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 3.G Current Proposed The licensee shall implement and maintain in effect Deleted. The licensee shall implement and all provisions of the approved fire protection maintain in effect all provisions of the approved fire program as described in the Updated Final Safety protection program as described in the Updated Analysis Report for the facility and as approved in Final Safety Analysis Report for the facility and as the Safety Evaluation Reports dated approved in the Safety Evaluation Reports dated March 22, 1978 with supplements dated December March 22, 1978 with supplements dated December 2, 1980, and February 12, 1981; January 19, 1983; 2, 1980, and February 12, 1981; January 19, 1983; July 17, 1987; September 28, 1987; and July 17, 1987; September 28, 1987; and January 5, 1989, subject to the following provision: January 5, 1989, subject to the following provision:

The licensee may make changes to the The licensee may make changes to the approved fire protection program without prior approved fire protection program without prior approval of the Commission only if those approval of the Commission only if those changes would not adversely affect the ability changes would not adversely affect the ability to to achieve and maintain safe shutdown in the achieve and maintain safe shutdown in the event of a fire. event of a fire.

Basis This license condition is proposed for deletion in its entirety consistent with the restriction of 10 CFR 50.82(a)(2) that Dresden, Unit 3, is no longer authorized to operate or emplace fuel in the reactor.

This license condition for making changes to the Fire Protection Program is no longer required to assure fire safety by maintaining the ability to achieve and maintain safe shutdown in the event of a fire.

License Condition 3.G, which is based on maintaining an operational fire protection program, in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shutdown of the reactors in the event of a fire, is no longer applicable for Dresden, Unit 3. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during plant decommissioning. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled plant facility is not required.

License Condition 3.N Current Proposed By Amendment No. 144, the license is amended to Deleted. By Amendment No. 144, the license is allow, on a one time temporary basis, operation of amended to allow, on a one time temporary basis, Dresden, Unit 3, with the corner room structural operation of Dresden, Unit 3, with the corner room steel members in the Low Pressure Coolant structural steel members in the Low Pressure Injection Corner Rooms outside the Updated Final Coolant Injection Corner Rooms outside the Safety Analysis Report (UFSAR) design Updated Final Safety Analysis Report (UFSAR) parameters. Operation under these conditions is design parameters. Operation under these allowed up to and including the next scheduled conditions is allowed up to and including the next refueling outage (D3R14). scheduled refueling outage (D3R14).

The repairs to Dresden, Unit 3, corner room The repairs to Dresden, Unit 3, corner room structural steel shall restore the steel design structural steel shall restore the steel design margins to the current UFSAR (updated through margins to the current UFSAR (updated through Revision 1A) design criteria. The design of the Revision 1A) design criteria. The design of the Page 32 of 85

Attachment 1 Evaluation of Proposed Changes modifications to the Dresden, Unit 3, corner room modifications to the Dresden, Unit 3, corner room structural steel members will be based on use of structural steel members will be based on use of elastic section modulus and the structural steel elastic section modulus and the structural steel stresses will be limited to 1.6 of the American stresses will be limited to 1.6 of the American Institute of Steel Construction (AISC allowables). Institute of Steel Construction (AISC allowables).

The modifications to Dresden, Unit 3, corner room The modifications to Dresden, Unit 3, corner room structural steel will be implemented during the structural steel will be implemented during the upcoming D3R14 refueling outage. upcoming D3R14 refueling outage.

During this interim period of operation, should During this interim period of operation, should vibratory ground motion exceeding the UFSAR vibratory ground motion exceeding the UFSAR Operating Basis Earthquake (OBE) design Operating Basis Earthquake (OBE) design parameters, Dresden, Unit 3, will be shut down for parameters, Dresden, Unit 3, will be shut down for inspection and will not start up without prior NRC inspection and will not start up without prior NRC approval. approval.

Basis This license condition is related to activities that have already occurred and are no longer applicable, and is proposed for deletion in its entirety. Additionally, once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 3.O - Additional Conditions Current Proposed The Additional Conditions contained in Appendix B, Deleted. The Additional Conditions contained in as revised through Amendment No. 185, are Appendix B, as revised through Amendment No.

hereby incorporated into this license. The licensee 185, are hereby incorporated into this license. The shall operate the facility in accordance with the licensee shall operate the facility in accordance with Additional Conditions. the Additional Conditions.

Basis This license condition is proposed for deletion in its entirety as the additional conditions in Appendix B to which it refers are proposed for deletion; therefore, retention of this license condition is unnecessary.

License Condition 3.V Current Proposed Exelon Generation Company, LLC shall relocate Deleted. Exelon Generation Company, LLC shall certain Technical Specification requirements to EGC relocate certain Technical Specification requirements controlled documents upon implementation of the to EGC controlled documents upon implementation Amendment No. 180. The items and appropriate of the Amendment No. 180. The items and documents are as described in Table LA, "Removal appropriate documents are as described in Table LA, of Details Matrix," and Table R, "Relocated "Removal of Details Matrix," and Table R, "Relocated Specifications," that are attached to the NRCs Specifications," that are attached to the NRCs Safety Evaluation enclosed with Amendment Safety Evaluation enclosed with Amendment No. 180. No. 180.

Basis This license condition resulted from the implementation of Improved TS for Dresden, Unit 3 (i.e., implementation of NUREG-1433), and is proposed for deletion in its entirety. It was related to the relocation of items from the TS to the Technical Requirements Manual. After certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, Unit 3, the 10 CFR 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant to Page 33 of 85

Attachment 1 Evaluation of Proposed Changes 10 CFR 50.82(a)(2); therefore, retention of these requirements are no longer necessary for the maintenance of the facility.

License Condition 3.W Current Proposed The schedule for performing Surveillance Deleted. The schedule for performing Surveillance Requirements (SRs) that are new or revised in Requirements (SRs) that are new or revised in Amendment No. 180 shall be as follows: Amendment No. 180 shall be as follows:

For SRs that are new in this amendment, For SRs that are new in this amendment, the first performance is due at the end of the first performance is due at the end of the first surveillance interval that begins on the first surveillance interval that begins on the date of implementation of Amendment the date of implementation of Amendment No. 180. No. 180.

For SRs that existed prior to this For SRs that existed prior to this amendment whose intervals of amendment whose intervals of performance are being reduced, the first performance are being reduced, the first reduced surveillance interval begins upon reduced surveillance interval begins upon completion of the first surveillance completion of the first surveillance performed after implementation of performed after implementation of Amendment No. 180. Amendment No. 180.

For SRs that existed prior to this For SRs that existed prior to this amendment that have modified acceptance amendment that have modified acceptance criteria, the first performance is due at the criteria, the first performance is due at the end of the first surveillance interval that end of the first surveillance interval that began on the date the surveillance was last began on the date the surveillance was last performed prior to the implementation of performed prior to the implementation of Amendment No. 180. Amendment No. 180.

For SRs that existed prior to this For SRs that existed prior to this amendment whose intervals of amendment whose intervals of performance are being extended, the first performance are being extended, the first extended surveillance interval begins upon extended surveillance interval begins upon completion of the last surveillance completion of the last surveillance performed prior to implementation of performed prior to implementation of Amendment No. 180. Amendment No. 180.

Basis This license condition resulted from the implementation of Improved TS for Dresden, Unit 3, (i.e., implementation of NUREG-1433) and is proposed for deletion in its entirety. Activities described in this license condition have already occurred, are no longer applicable; therefore, the maintenance of these requirements is no longer necessary. Additionally, once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to activities related to the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

License Condition 3.X Current Proposed The license is amended to authorize changing the Deleted. The license is amended to authorize UFSAR to allow credit for containment changing the UFSAR to allow credit for containment Page 34 of 85

Attachment 1 Evaluation of Proposed Changes overpressure as detailed below, to assure adequate overpressure as detailed below, to assure adequate Net Positive Suction Head is available for low Net Positive Suction Head is available for low pressure Emergency Core Cooling System pumps pressure Emergency Core Cooling System pumps following a design basis accident. following a design basis accident.

From (sec) To (sec) Credit (psig) From (sec) To (sec) Credit (psig)

Accident start 290 9.5 Accident start 290 9.5 290 5,000 4.8 290 5,000 4.8 5,000 30,000 6.6 5,000 30,000 6.6 30,000 40,000 6.0 30,000 40,000 6.0 40,000 45,500 5.4 40,000 45,500 5.4 45,500 52,500 4.9 45,500 52,500 4.9 52,500 60,500 4.4 52,500 60,500 4.4 60,500 70,000 3.8 60,500 70,000 3.8 70,000 84,000 3.2 70,000 84,000 3.2 84,000 104,000 2.5 84,000 104,000 2.5 104,000 136,000 1.8 104,000 136,000 1.8 136,000 Accident end 1.1 136,000 Accident end 1.1 Basis This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been removed from the reactor, 10 CFR 50.82(a)(2) prohibits operation of the Dresden, Unit 3, reactor. With operation of the reactor prohibited, the Emergency Core Cooling System pumps will no longer be required for mitigation of design basis accidents and as such, the crediting of containment overpressure allowed by License Condition 3.X is no longer needed.

License Condition 3.Y - Updated Final Safety Analysis Report Current Proposed The Exelon Generation Company, LLC Updated Deleted. The Exelon Generation Company, LLC Final Safety Analysis Report supplement, submitted Updated Final Safety Analysis Report supplement, pursuant to 10 CFR 54.21(d), describes certain submitted pursuant to 10 CFR 54.21(d), describes future activities to be completed prior to the period certain future activities to be completed prior to the of extended operation. The Exelon Generation period of extended operation. The Exelon Company, LLC shall complete these activities no Generation Company, LLC shall complete these later than January 12, 2011, and shall notify the activities no later than January 12, 2011, and shall NRC in writing when implementation of these notify the NRC in writing when implementation of activities is complete and can be verified by NRC these activities is complete and can be verified by inspection. NRC inspection.

The Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement, as revised, shall be included in the supplement, as revised, shall be included in the next next scheduled update to the Updated Final Safety scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e}(4} Analysis Report required by 10 CFR 50.71 (e}(4}

following issuance of this renewed license. Until following issuance of this renewed license. Until that that update is complete, Exelon Generation update is complete, Exelon Generation Company, Company, LLC may make changes to the programs LLC may make changes to the programs and and activities described in the supplement without activities described in the supplement without prior prior Commission approval, provided that Exelon Commission approval, provided that Exelon Generation Company, LLC evaluates such change Generation Company, LLC evaluates such change pursuant to the criteria set forth in 10 CFR 50.59 pursuant to the criteria set forth in 10 CFR 50.59 and Page 35 of 85

Attachment 1 Evaluation of Proposed Changes and otherwise complies with the requirements in otherwise complies with the requirements in that that section. section.

Basis This license condition imposed program requirements necessary for the Dresden, Unit 3, to continue operation beyond January 12, 2011. After certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, Unit 2, the 10 CFR 50 license will no longer authorize operation of the reactor pursuant to 10 CFR 50.82(a)(2). Thus, the program requirements to support extended operation are not necessary.

License Condition 3.Z Current Proposed All capsules in the reactor vessel that are removed Deleted. All capsules in the reactor vessel that are and tested must meet the test procedures and removed and tested must meet the test procedures reporting requirements of ASTM E 185 82 to the and reporting requirements of ASTM E 185 82 to the extent practicable for the configuration of the extent practicable for the configuration of the specimens in the capsule. Any changes to the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to capsules, must be approved by the NRC prior to implementation. All capsules placed in storage implementation. All capsules placed in storage must must be maintained for future insertion. be maintained for future insertion.

Basis This license condition was issued concurrent with the Renewed Facility Operating License on October 28, 2004. This license condition is described in NUREG-1796, "Safety Evaluation Report Related to the License Renewal of Dresden Nuclear Power Station, Units 2 and 3," Section 1.7, "Summary of Proposed License Conditions," of issued October 2004 (Reference 6).

This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility. Continued implementation of the applicable surveillance capsule testing and reporting requirements are no longer necessary for Dresden, Unit 3, because: (a) Exelon has decided to cease power operations of Dresden, Unit 3, and (b) from a fracture toughness perspective, the Dresden, Unit 3, RPV will cease to be exposed to further irradiation by high energy neutrons or subjected to any high thermal stress environments, as induced by operating the RCS at an elevated temperature. Further, 10 CFR 50.60(a) stipulates that reactor facilities for which the certifications required under § 50.82(a)(1) have been submitted, no longer need to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in Appendices G and H.

The physical and radiological control of the remaining surveillance capsules that are located in the RPV will be managed in accordance with the applicable radiological control requirements of 10 CFR Part 20 and with any applicable security or physical protection requirements for components in either 10 CFR Part 37 or 10 CFR Part 73. Therefore, the removal, testing, reporting, and storage requirements for reactor vessel surveillance capsules and their test specimens do not need to be implemented further once Dresden, Unit 3, permanently ceases power operations. There will no longer be any need to remove the remaining surveillance capsules from the RPV or perform material testing of the test specimens in those capsules. As such, deletion of this license condition is appropriate. Any corresponding commitments in the Dresden, Units 2 and 3, UFSAR will also be deleted under the provisions of 10 CFR 50.59 upon NRC approval of this proposed license amendment request.

Page 36 of 85

Attachment 1 Evaluation of Proposed Changes License Condition 3.CC Current Proposed Upon implementation of Amendment No. 218 Deleted. Upon implementation of Amendment No.

adopting TSTF-448, Revision 3, the determination 218 adopting TSTF-448, Revision 3, the of control room envelope (CRE) unfiltered air determination of control room envelope (CRE) inleakage as required by SR 3.7.4.4, in accordance unfiltered air inleakage as required by SR 3.7.4.4, with TS 5.5.14.c.(i), the assessment of CRE in accordance with TS 5.5.14.c.(i), the assessment habitability as required by Specification 5.5.14.c.(ii), of CRE habitability as required by Specification and the measurement of CRE pressure as required 5.5.14.c.(ii), and the measurement of CRE pressure by Specification 5.5.14.d, shall be considered met. as required by Specification 5.5.14.d, shall be Following implementation: considered met. Following implementation:

(1) The first performance of SR 3.7.4.4, in (1) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), accordance with Specification 5.5.14.c.(i),

shall be within the specified Frequency of shall be within the specified Frequency of 6 years, plus the 18-month allowance of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from January 1997, SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer the date of the most recent successful tracer gas test, as stated in the December 9, 2003 gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or letter response to Generic Letter 2003-01, or within the next 18 months if the time period within the next 18 months if the time period since the most recent successful tracer gas since the most recent successful tracer gas test is greater than 6 years. test is greater than 6 years.

(2) The first performance of the periodic (2) The first performance of the periodic assessment of CRE habitability, Specification assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured 9-month allowance of SR 3.0.2, as measured from January 1997, the date of the most from January 1997, the date of the most recent successful tracer gas test, as stated in recent successful tracer gas test, as stated in the December 9, 2003 letter response to the December 9, 2003 letter response to Generic Letter 2003-01, or within the next Generic Letter 2003-01, or within the next 9 months if the time period since the most 9 months if the time period since the most recent successful tracer gas test is greater recent successful tracer gas test is greater than 3 years. than 3 years.

(3) The first performance of the periodic (3) The first performance of the periodic measurement of CRE pressure, Specification measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, as measured 6 months allowed by SR 3.0.2, as measured from the date of the most recent successful from the date of the most recent successful pressure measurement test, or within pressure measurement test, or within 6 months if not performed previously. 6 months if not performed previously.

Basis This license condition is proposed for deletion in its entirety. The proposed change removes the requirements of TSTF-448 that involve assessing the Control Room Envelope (CRE) Habitability at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. These assessments were completed in accordance with the schedule specified in the license condition.

Page 37 of 85

Attachment 1 Evaluation of Proposed Changes Exelon performed an analysis for an FHA in the SFP for dose results for the control room (CR) after permanent shutdown. The calculation accounts for radioactive material inventory in the most recently irradiated elements in the SFP after 48 days of decay following permanent shutdown. For the analysis, Exelon took no credit for filtered recirculation of the control room air. The results of the calculation showed that the dose consequences to occupants in the control room were below the 10 CFR 50.67 dose limit of 5 rem. Based on the fact that the dose at the CR is less than the 10 CFR 50.67 dose limit and that no credit was taken for CR filtered recirculation, the CRE Habitability Program is not required to provide airborne radiological protection for the control room operators. This submittal also proposes to remove TS 3.7.4 for the control room emergency ventilation system and TS 5.5.14 for the CRE Habitability Program.

Since TS 3.7.4 and TS 5.5.14 are no longer necessary, this license condition is no longer needed; therefore, it is proposed for deletion.

Proposed License Condition 3.EE Current Proposed Handling of irradiated fuel in the spent fuel pool

[None] will not be permitted following implementation of the PDTS until a minimum of 48 days following permanent shutdown.

Basis Exelon is proposing a new license condition such that initial system abandonment activities may be started expeditiously after the permanent removal of fuel from the reactor vessel. By applying this new license condition, Exelon will be able to remove the TS requirements associated with those systems that perform mitigative actions assumed in the FHA analyses by precluding the possibility of an FHA until after the 48-day decay period assumed in the post permanent shutdown FHA has elapsed.

Once the reactor has been permanently defueled with all irradiated fuel placed in the SFP and the certifications submitted in accordance with 10 CFR 50.82, power operation or emplacement of fuel in the reactor will not be allowed. Therefore, all DBAs associated with power operations or fuel handling in the reactor will no longer be applicable, which provides the basis for removal of the Safety Limits and most of the Limiting Conditions for Operation.

In order to implement the PDTS prior to the 48-day decay time assumed in the post permanent shutdown FHA analysis, Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 days after permanent shutdown through the imposition of the proposed license condition. This effectively prevents an FHA from occurring until after the 48-day decay period has elapsed.

License Condition 4 Current Proposed This renewed operating license is effective as of the This renewed operating license is effective as of the date of issuance and shall expire at midnight on date of issuance and shall expire at midnight on January 12, 2031. January 12, 2031 is effective until the Commission notifies the licensee in writing that the license is terminated.

Basis This License Condition is revised to reflect the permanently defueled condition of the facility. Once Dresden has permanently ceased operation and Exelon has certified that fuel has been removed from the reactors, 10 CFR 50.82(a)(2) prohibits operation of the Dresden reactors. Reference to an operating state Page 38 of 85

Attachment 1 Evaluation of Proposed Changes for the facility is deleted and this license condition is proposed for further revision in accordance with 10 CFR 50.51(b) in that the license authorizes ownership and possession by Exelon until the Commission notifies the licensee in writing that the license is terminated.

Appendix B - Additional Conditions - Facility Operating License No. DPR-25 Current Proposed Deleted Basis This Appendix to the RFOL contains additional conditions that are related to activities that have already occurred and are proposed for deletion in their entirety. Once Dresden, Unit 3, has permanently ceased operation and certified that fuel has been permanently removed from the reactor, reference to these activities that are associated with the operation of the facility would be inconsistent with the provisions of 10 CFR 50.82(a)(2).

Page 39 of 85

Attachment 1 Evaluation of Proposed Changes Detailed Description of the Proposed Changes to the Dresden TS The following tables provides a summary describing which Dresden, Units 2 and 3, TS are being deleted in their entirety and which TS are being retained into the PDTS. The details and justification for the proposed changes are provided, arranged by TS section.

TS Section 1.0 - Use and Application TS Being Deleted TS Being Retained 1.1 - Definitions 1.2 - Logical Connectors 1.3 - Completion Times 1.4 - Frequency Units 2 and 3, TS Section 1.1 - Definitions TS Section 1.1, "Definitions," provides defined terms that are applicable throughout the TS and TS Bases.

A number of definitions are being proposed to be deleted because they have no relevance to, and no longer apply to the permanently defueled facility status.

Definition Basis for Change AVERAGE PLANAR LINEAR HEAT This definition is proposed for deletion in the PDTS since the term is GENERATION RATE (APLHGR) not used in any PDTS specification. This term is only meaningful to a reactor authorized to operate.

CERTIFIED FUEL HANDLER A new definition for Certified Fuel Handler was proposed for addition to the PDTS in the LAR proposing changes to TS Sections 1.1 and 5.0 (Reference 2) which is currently under NRC review.

CHANNEL CALIBRATION This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

CHANNEL CHECK This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

CHANNEL FUNCTIONAL TEST This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

CORE ALTERATION This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term has no meaning when there is no fuel in the reactor vessels.

CORE OPERATING LIMITS This definition is proposed for deletion in the PDTS since the term is REPORT (COLR) not used in any PDTS specification, and Specification 5.6.5 that requires the COLR is also proposed for elimination.

DOSE EQUIVALENT I-131 This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term is used to express dose from a mixture of iodine isotopes created in an operating core and contained in plant primary or secondary coolant. The value of Dose Equivalent I-131 is used for dose analysis of accidents involving primary coolant releases. Those accident conditions will no longer apply to the permanently shutdown and defueled facility.

DRAIN TIME This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term is no longer applicable since fuel will be permanently removed from the reactors.

Page 40 of 85

Attachment 1 Evaluation of Proposed Changes INSERVICE TESTING PROGRAM This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. Inservice testing in accordance with 10 CFR 50.55a will no longer be required once the reactors are permanently shutdown and defueled.

LEAKAGE This definition is proposed for deletion in the PDTS since because none of the structures, systems, or components (SSCs) from or into which leakage is monitored are credited in the analysis of an FHA the remaining credible accidents.

LINEAR HEAT GENERATION This definition is proposed for deletion in the PDTS since the term is RATE (LHGR) not used in any PDTS specification. This term is only meaningful to a reactor authorized to operate.

LOGIC SYSTEM FUNCTIONAL This definition is proposed for deletion in the PDTS since the term is TEST not used in any PDTS specification. There are no logic systems credited in the analysis of the accidents that remain credible in the permanently defueled condition.

MINIMUM CRITICAL POWER This definition is proposed for deletion in the PDTS since the term is RATIO (MCPR) not used in any PDTS specification. This term is only meaningful to a reactor authorized for critical operation.

MODE This definition, including TS Table 1.1-1, is proposed for deletion in the PDTS since operating Modes are not used in any PDTS In conjunction with deletion of the specification. Modes are defined for operating or refueling term "Mode," TS Table 1.1-1, conditions. This term does not apply to a facility in the permanently "MODES," is also being deleted. defueled condition.

NON-CERTIFIED OPERATOR A new definition for Non-Certified Operator is proposed for addition to the PDTS in the LAR proposing changes to Units 2 and 3, TS Sections 1.1 and 5.0 (Reference 3) which is currently under NRC review.

OPERABLE - OPERABILITY This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification.

RATED THERMAL POWER (RTP) This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term is only meaningful to a reactor authorized to operate.

REACTOR PROTECTION SYSTEM This definition is proposed for deletion in the PDTS since the term is (RPS) RESPONSE TIME not used in any PDTS specification. The RPS will have no function for the permanently shutdown and defueled facility.

SHUTDOWN MARGIN (SDM) This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term is only meaningful to reactors authorized to operate.

THERMAL POWER This definition is proposed for deletion in the PDTS since the term is not used in any PDTS specification. This term is only meaningful to reactors authorized to operate.

TURBINE BYPASS SYSTEM This definition is proposed for deletion in the PDTS since the term is RESPONSE TIME not used in any PDTS specification. The main turbine has no function in the permanently defueled condition.

Basis With the exception of the two new definitions, all of the definitions in the above table are proposed for deletion since they are relevant to an operating reactor and are not used in the PDTS. 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)(i) and (ii), 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. With the submittal of the certifications in accordance with Page 41 of 85

Attachment 1 Evaluation of Proposed Changes 10 CFR 50.82(a)(1), Exelon will no longer be authorized to operate the facility; therefore, the referenced definitions will no longer be relevant.

Units 2 and 3, TS Section 1.2, Logical Connectors TS 1.2, "Logical Connectors," contains an explanation of the logical connectors used to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies throughout TS. Logical Connectors are no longer used in the PDTS. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, Units 2 and 3, 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). Dresden, Units 2 and 3, will have irradiated fuel in the SFPs and, after a 48-day decay period, the only required LCO will be 3.7.8, Spent Fuel Storage Pool Water Level. This LCO has a single Condition prompting a single Action and does not use logical connectors, which obviates the need for the explanatory text and examples in Section 1.2.

TS Section 1.3 - Completion Times TS 1.3, "Completion Times," establishes the Completion Time convention and provides guidance for its use. It is modified to reflect the permanently shutdown and defueled condition and the Completion Times that continue to exist in the PDTS.

Current Background Proposed Background Limiting Conditions for Operation (LCOs) specify Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation minimum requirements for ensuring safe of the unit. The ACTIONS associated with an LCO operation of the unitstorage and handling of state Conditions that typically describe the ways in irradiated fuel. The ACTIONS associated with which the requirements of the LCO can fail to be an LCO state Conditions that typically describe met. Specified with each stated Condition are the ways in which the requirements of the LCO Required Action(s) and Completion Time(s). can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

Current Description Proposed Description The Completion Time is the amount of time allowed The Completion Time is the amount of time for completing a Required Action. It is referenced to allowed for completing a Required Action. It is the discovery of a situation (e.g., inoperable referenced to the discovery of a situation (e.g.,

equipment or variable not within limits) that requires inoperable equipment or variable not within limits) entering an ACTIONS Condition unless otherwise that requires entering an ACTIONS Condition specified, providing the unit is in a MODE or unless otherwise specified, providing the specified condition stated in the Applicability of the unitfacility is in a MODE or specified condition LCO. Unless otherwise specified, the Completion stated in the Applicability of the LCO. Unless Time begins when a senior licensed operator on the otherwise specified, the Completion Time begins operating shift crew with responsibility for plant when a senior licensed operator on the operating operations makes the determination that an LCO is shift crew with responsibility for plant operations not met and an ACTIONS Condition is entered. The makes the determination that an LCO is not met "otherwise specified" exceptions are varied, such as and an ACTIONS Condition is entered. The a Required Action Note or Surveillance "otherwise specified" exceptions are varied, such Requirement Note that provides an alternative time as a Required Action Note or Surveillance to perform specific tasks, such as testing, without Requirement Note that provides an alternative starting the Completion Time. While utilizing the time to perform specific tasks, such as testing, Note, should a Condition be applicable for any without starting the Completion Time. While reason not addressed by the Note, the Completion utilizing the Note, should a Condition be Page 42 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 1.3 - Completion Times Time begins. Should the time allowance in the Note applicable for any reason not addressed by the be exceeded, the Completion Time begins at that Note, the Completion Time begins. Should the point. The exceptions may also be incorporated into time allowance in the Note be exceeded, the the Completion Time. For example, LCO 3.8.1, "AC Completion Time begins at that point. The Sources - Operating," Required Action B.2, requires exceptions may also be incorporated into the declaring required feature(s) supported by an Completion Time. For example, LCO 3.8.1, "AC inoperable diesel generator, inoperable when the Sources - Operating," Required Action B.2, redundant required feature(s) are inoperable. The requires declaring required feature(s) supported Completion Time states, "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery of by an inoperable diesel generator, inoperable Condition B concurrent with inoperability of when the redundant required feature(s) are redundant required feature(s)." In this case the inoperable. The Completion Time states, "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time does not begin until the conditions from discovery of Condition B concurrent with in the Completion Time are satisfied. Required inoperability of redundant required feature(s)." In Actions must be completed prior to the expiration of this case the Completion Time does not begin the specified Completion Time. An ACTIONS until the conditions in the Completion Time are Condition remains in effect and the Required satisfied. Required Actions must be completed Actions apply until the Condition no longer exists or prior to the expiration of the specified Completion the unit is not within the LCO Applicability. Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility unit is not within the LCO Applicability.

If situations are discovered that require entry into If situations are discovered that require entry into more than one Condition at a time within a single more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for LCO (multiple Conditions), the Required Actions each Condition must be performed within the for each Condition must be performed within the associated Completion Time. When in multiple associated Completion Time. When in multiple Conditions, separate Completion Times are tracked Conditions, separate Completion Times are for each Condition starting from the discovery of the tracked for each Condition starting from the situation that required entry into the Condition, discovery of the situation that required entry into unless otherwise specified. the Condition, unless otherwise specified.

Once a Condition has been entered, subsequent Once a Condition has been entered, subsequent divisions, subsystems, components, or variables divisions, subsystems, components, or variables expressed in the Condition, discovered to be expressed in the Condition, discovered to be inoperable or not within limits, will not result in inoperable or not within limits, will not result in separate entry into the Condition unless specifically separate entry into the Condition unless stated. The Required Actions of the Condition specifically stated. The Required Actions of the continue to apply to each additional failure, with Condition continue to apply to each additional Completion Times based on initial entry into the failure, with Completion Times based on initial Condition, unless otherwise specified. entry into the Condition, unless otherwise However, when a subsequent division, subsystem, specified.

component, or variable expressed in the Condition However, when a subsequent division, is discovered to be inoperable or not within limits, subsystem, component, or variable expressed in the Completion Time(s) may be extended. To apply the Condition is discovered to be inoperable or not this Completion Time extension, two criteria must within limits, the Completion Time(s) may be first be met. The subsequent inoperability: extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

Page 43 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 1.3 - Completion Times

a. Must exist concurrent with the first inoperability; a. Must exist concurrent with the first and inoperability; and

b. Must remain inoperable or not within limits after b. Must remain inoperable or not within limits the first inoperability is resolved. after the first inoperability is resolved.

The total Completion Time allowed for completing a The total Completion Time allowed for completing Required Action to address the subsequent a Required Action to address the subsequent inoperability shall be limited to the more restrictive inoperability shall be limited to the more restrictive of either: of either:

a. The stated Completion Time, as measured a. The stated Completion Time, as from the initial entry into the Condition, plus an measured from the initial entry into the Condition, additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or

b. The stated Completion Time as measured from b. The stated Completion Time as measured discovery of the subsequent inoperability. from discovery of the subsequent inoperability.

The above Completion Time extension does not The above Completion Time extension does not apply to those Specifications that have exceptions apply to those Specifications that have exceptions that allow completely separate re-entry into the that allow completely separate re-entry into the Condition (for each division, subsystem, component Condition (for each division, subsystem, or variable expressed in the Condition) and component or variable expressed in the separate tracking of Completion Times based on Condition) and separate tracking of Completion this re-entry. These exceptions are stated in Times based on this re-entry. These exceptions individual Specifications. are stated in individual Specifications.

The above Completion Time extension does not The above Completion Time extension does not apply to a Completion Time with a modified "time apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed zero." This modified "time zero" may be as a repetitive time (i.e., "once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," where expressed as a repetitive time (i.e., "once per 8 the Completion Time is referenced from a previous hours," where the Completion Time is referenced completion of the Required Action versus the time of from a previous completion of the Required Action Condition entry) or as a time modified by the phrase versus the time of Condition entry) or as a time "from discovery . . ." modified by the phrase "from discovery . . ."

Basis The Background section of TS 1.3 is modified to reflect the upcoming change in status regarding Dresden. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). As a result, the primary mission will change from the safe operation of the units to the safe storage and handling of irradiated fuel.

The Description section of TS 1.3 is modified to reflect the change in status regarding Dresden. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). As a result, the PDTS will contain no operability requirements for any equipment. In addition, the term facility better represents Dresden in the permanently shutdown and defueled condition.

Page 44 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 1.3 - Completion Times The examples in this section are proposed for deletion. The examples are no longer necessary because they describe the Completion Times that do not remain in the PDTS. The Action that remains in the PDTS must be completed "IMMEDIATELY."

TS Section 1.4 - Frequency TS 1.4, "Frequency," defines the proper use and application of Frequency requirements. It is modified to reflect the permanently shutdown and defueled condition and the Frequency that continues to exist in the PDTS.

Current Description Proposed Description The "specified Frequency" is referred to The "specified Frequency" is referred to throughout this section and each of the throughout this section and each of the Specifications of Section 3.0, Surveillance Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency" consists of the requirements of the Frequency column of each SR as well as certain Frequency column of each SR as well as certain Notes in the Surveillance column that modify Notes in the Surveillance column that modify performance requirements. performance requirements.

Sometimes special situations dictate when the Sometimes special situations dictate when the requirements of a Surveillance are to be met. They requirements of a Surveillance are to be met.

are "otherwise stated" conditions allowed by SR They are "otherwise stated" conditions allowed by 3.0.1. They may be stated as clarifying Notes in the SR 3.0.1. They may be stated as clarifying Notes Surveillance, as part of the Surveillance, or both. in the Surveillance, as part of the Surveillance, or Example 1.4-4 discusses these special situations. both. Example 1.4-4 discusses these special situations.

Situations where a Surveillance could be required Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not (i.e., its Frequency could expire), but where it is possible or not desired that it be performed until not possible or not desired that it be performed sometime after the associated LCO is within its until sometime after the associated LCO is within Applicability, represent potential SR 3.0.4 conflicts. its Applicability, represent potential SR 3.0.4 To avoid these conflicts, the SR (i.e., the conflicts. To avoid these conflicts, the SR (i.e.,

Surveillance or the Frequency) is stated such that it the Surveillance or the Frequency) is stated such is only "required" when it can be and should be that it is only "required" when it can be and should performed. With an SR satisfied, SR 3.0.4 imposes be performed. With an SR satisfied, SR 3.0.4 no restriction. imposes no restriction.

The use of "met" or "performed" in these instances The use of "met" or "performed" in these conveys special meaning. A surveillance is "met" instances conveys special meaning. A only when the acceptance criteria are satisfied. surveillance is "met" only when the acceptance Known failure of the requirement of a Surveillance, criteria are satisfied. Known failure of the even without a Surveillance specifically being requirement of a Surveillance, even without a "performed," constitutes a Surveillance not "met." Surveillance specifically being "performed,"

"Performance" refers only to the requirement to constitutes a Surveillance not "met."

specifically determine the ability to meet the "Performance" refers only to the requirement to acceptance criteria. SR 3.0.4 restrictions would not specifically determine the ability to meet the apply if both the following conditions are satisfied: acceptance criteria. SR 3.0.4 restrictions would Page 45 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 1.4 - Frequency

a. The Surveillance is not required to be not apply if both the following conditions are performed; and satisfied:

b. The Surveillance is not required to be met a. The Surveillance is not required to be or, even if required to be met, is not known performed; and to be failed.

b. The Surveillance is not required to be met or, even if required to be met, is not known to be failed.

Current Examples Proposed Example The following examples illustrate the various ways The following examples illustrates the manner in that Frequencies are specified. In these examples, which various ways that Frequencies are the Applicability of the LCO (LCO not shown) is specified. In these examples, the Applicability of MODES 1, 2, and 3. The examples do not reflect the LCO (LCO not shown) is MODES 1, 2, and 3.

the potential application of LCO 3.0.4.b. The examples do not reflect the potential application of LCO 3.0.4.b.

Example 1.4-1 Example 1.4-1 is modified to address an example of the Frequency that is utilized by TS 3.7.8.

Example 1.4-2 Example 1.4-2 is proposed for deletion.

Example 1.4-3 Example 1.4-3 is proposed for deletion.

Example 1.4-4 Example 1.4-4 is proposed for deletion.

Basis TS Section 1.4 is modified to reflect the change in status regarding Dresden. This includes modifications to the description section and to the examples. These proposed changes are editorial changes that reflect the changes to the other TS and the remaining requirements.

After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). As a result, the number and types of Surveillance Requirements (SRs) that remain the TS are limited to one SR in Technical Specification S 3.7.8. This section is modified to provide the rules of usage and an example that continues to be applicable for that technical specification.

Example 1.4-1 is modified to address an example of a Frequency that is utilized by TS 3.7.8. This includes the elimination of the references to the term "operational," inoperable equipment, Modes, Examples 1.4-3 and 1.4-4, and LCO 3.0.4, and replacing the term "unit" with "facility."

Examples 1.4-2, 1.4-3, and 1.4-4 are proposed for deletion.

TS Section 2.0, Safety Limits (SLs)

TS Being Deleted TS Being Retained 2.1 Safety Limits (SLs) 2.2 SL Violations Basis Page 46 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 2.0, Safety Limits (SLs), contains limits upon important process variables to assure the integrity of the fuel cladding and the RCS in all modes of operation. Pursuant to 10 CFR 50.36(c)(1),

safety limits are limiting parameters necessary to protect the physical barriers that guard against uncontrolled release of radioactivity from a nuclear reactor. The safety limits and safety limit violations TS apply to the reactor core and the RCS; they have no function in the permanently defueled condition.

These specifications do not apply to the safe storage and handling of irradiated fuel in the SFPs.

Summary TS 2.0 is being proposed for deletion in its entirety since the safety limits do not apply to reactors that are in a permanently defueled condition. Once Exelon submits the certifications required by 10 CFR 50.82(a)(1) for Dresden, 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).

These specifications do not apply to the safe storage and handling of irradiated fuel in the SFPs. With TS Section 2.0 deleted in its entirety, the applicable Bases will also be deleted to reflect this change.

TS Section 3.0 - Limiting Condition for Operation (LCO) Applicability TS Being Deleted TS Being Retained 3.0 Limiting Condition for Operation (LCO)

Applicability 3.0 Surveillance Requirement (SR) Applicability TS Section 3.0, "Limiting Condition for Operation (LCO) Applicability," establishes the general requirements for all Specifications and applies at all times, unless otherwise stated. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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).

Consequently, some LCOs are no longer applicable and some must be revised to reflect the permanently defueled condition.

TS Section 3.0, "Surveillance Requirement (SR) Applicability," establishes the general requirements for all Specifications in Sections 3.1 through 3.10 and applies at all times, unless otherwise stated. SR 3.0.2 and SR 3.0.3 apply in Chapter 5 only when invoked by a Chapter 5 Specification. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). Consequently, some SRs must be revised to reflect the permanently defueled condition.

Current LCO 3.0.1 Proposed LCO 3.0.1 LCOs shall be met during the MODES or other LCOs shall be met during the MODES or other specified conditions in the Applicability, except as specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, and LCO 3.0.8. provided in LCO 3.0.2, LCO 3.0.7, and LCO 3.0.8.

Current LCO 3.0.2 Proposed LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and be met, except as provided in LCO 3.0.5 and LCO LCO 3.0.6. 3.0.6.

If the LCO is met or is no longer applicable prior to If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), expiration of the specified Completion Time(s),

Page 47 of 85

Attachment 1 Evaluation of Proposed Changes completion of the Required Action(s) is not required, completion of the Required Action(s) is not required, unless otherwise stated. unless otherwise stated.

Additional Current LCOs in TS Section 3.0 Additional Proposed LCOs in TS Section 3.0 LCO 3.0.3 LCO 3.0.3 is proposed for deletion.

LCO 3.0.4 LCO 3.0.4 is proposed for deletion.

LCO 3.0.5 LCO 3.0.5 is proposed for deletion.

LCO 3.0.6 LCO 3.0.6 is proposed for deletion.

LCO 3.0.7 LCO 3.0.7 is proposed for deletion.

LCO 3.0.8 LCO 3.0.8 is proposed for deletion.

Current LCO 3.0.9 Proposed LCO 3.0.9 LCOs, including associated ACTIONS, shall apply LCOs, including associated ACTIONS, shall apply to each unit individually, unless otherwise indicated. to each unit spent fuel storage pool individually, Whenever the LCO refers to a system or component unless otherwise indicated. Whenever the LCO that is shared by both units, the ACTIONS will apply refers to a system or component that is shared by to both units simultaneously. both units, the ACTIONS will apply to both units simultaneously.

Basis LCO 3.0.1 is modified by eliminating the references to MODES because this term does not apply to a facility in the permanently defueled condition. MODES as defined in Table 1.1-1 are defined for operating or refueling conditions. Table 1.1-1 was proposed for deletion in TS Section 1.0. In addition, the references to LCOs 3.0.7 and 3.0.9 are deleted to reflect the proposed deletion of those LCOs discussed below.

LCO 3.0.2 is modified by eliminating the references to LCOs 3.0.5 and 3.0.6. This change reflects the proposed deletion of those LCOs as discussed below. Additionally, the second paragraph of this LCO that references the possibility of an LCO being met prior to the expiration of the Completion Time(s) is deleted as only one Required Action is proposed for inclusion in the PDTS. The remaining Required Action will have a Completion Time of "Immediately;" therefore, this paragraph is not needed.

LCO 3.0.3 provides the actions that must be implemented when an LCO is not met. It is only applicable in MODES 1, 2, and 3. Pursuant to 10 CFR 50.82(a)(2), the facility licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. Therefore, reference to operational MODES is no longer relevant, and LCO 3.0.3 is no longer applicable in the permanently defueled condition.

LCO 3.0.4 provides limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. LCO 3.0.4 is not proposed for inclusion in the PDTS since all actions in the remaining Technical Specification that has a Required Action is 3.7.8, which contains only one Required Action with a Completion Time of "Immediately." This makes LCO 3.0.4 unnecessary. Thus, LCO 3.0.4 is no longer applicable in the permanently defueled condition.

LCO 3.0.5 provides the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The allowance of LCO 3.0.5 to not comply with the requirements of LCO 3.0.2 (i.e., to not comply with the Required Actions)

Page 48 of 85

Attachment 1 Evaluation of Proposed Changes to allow the performance of SRs on equipment declared inoperable or removed from service is no longer required. The remaining PDTS ACTION does not include requirements to declare equipment inoperable or to remove it from service.

LCO 3.0.6 addresses the actions required for a supported system when the support system LCO is not met. It is proposed for deletion since there are no LCOs that contain equipment that must be operable or in operation in the PDTS.

LCO 3.0.7 pertains to certain special tests and operations required to be performed at various times over the life of the unit. It is proposed for deletion since special tests and operations are not applicable to a permanently defueled facility.

LCO 3.0.8 addresses the actions required when one or more required snubbers are unable to perform their associated support function(s). It is proposed for deletion because there are no LCOs that contain equipment that must be operable or in operation in the PDTS. Snubbers are not required to perform any TS function.

LCO 3.0.9 addresses the application of TS ACTIONS to the individual units and systems or components shared by both units. It is proposed for revision since unit specific TS ACTIONS will only be applicable to the SFPs in the permanently defueled facility.

TS Section 3.0 - Surveillance Requirement (SR) Applicability Current SR 3.0.1 Proposed SR 3.0.1 SRs shall be met during the MODES or other SRs shall be met during the MODES or other specified condition in the Applicability for individual specified condition in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is meet a Surveillance, whether such failure is experienced during the performance of the experienced during the performance of the Surveillance or between performances of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Surveillance, shall be failure to meet the LCO.

Failure to perform a Surveillance within the specified Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or to be performed on inoperable equipment or variables outside specified limits. variables outside specified limits.

Current SR 3.0.2 Proposed SR 3.0.2 The specified Frequency for each SR is met if the The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured interval specified in the Frequency, as measured from the previous performance or as measured from from the previous performance or as measured from the time a specified condition of the Frequency is the time a specified condition of the Frequency is met. met.

For Frequencies specified as "once," the above For Frequencies specified as "once," the above interval extension does not apply. interval extension does not apply.

If a Completion Time requires periodic performance If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency on a "once per . . ." basis, the above Frequency extension applies to each performance after the extension applies to each performance after the initial performance. initial performance.

Page 49 of 85

Attachment 1 Evaluation of Proposed Changes Exceptions to this Specification are stated in the Exceptions to this Specification are stated in the individual Specifications. individual Specifications.

Current SR 3.0.4 Proposed SR 3.0.4 Entry into a MODE or other specified condition in Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their the LCO's Surveillances have been met within their specified Frequency, except as provided by SR specified Frequency, except as provided by 3.0.3. When an LCO is not met due to SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a Surveillances not having been met, entry into a MODE or other specified condition in the MODE or other specified condition in the Applicability shall only be made in accordance with Applicability shall only be made in accordance with LCO 3.0.4. LCO 3.0.4.

This provision shall not prevent entry into MODES This provision shall not prevent entry into MODES or other specified conditions in the Applicability that or other specified conditions in the Applicability that are required to comply with ACTIONS or that are are required to comply with ACTIONS or that are part of a shutdown of the unit. part of a shutdown of the unit.

Current SR 3.0.5 Proposed SR 3.0.5 SRs shall apply to each unit individually, unless SRs shall apply to each unit spent fuel storage otherwise indicated. pool individually, unless otherwise indicated.

Basis SR 3.0.1 is modified by deleting the references to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. MODES, defined in Table 1.1-1 for operation or refueling conditions, are not used in any PDTS specification. In addition, SR 3.0.1 is modified by eliminating the discussion regarding inoperable equipment. The remaining LCOs do not include any equipment operability requirements.

SR 3.0.2 provides an allowance for extending the frequency for performance of a SR to 1.25 times the interval specified in the frequency to facility scheduling or unforeseen problems that may prevent performance during normal intervals. It is proposed for revision to remove conditions for frequencies that do not exist in PDTS.

SR 3.0.4 is modified by deleting the references to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. MODES, defined in Table 1.1-1 for operation or refueling conditions, are not used in any PDTS specification. SR 3.0.4 is modified by eliminating the provision that states that it shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with Actions or that are part of a shutdown of the unit. The only remaining Technical Specification with a Required Actions is TS 3.7.8, and it does not contain any Required Actions that would require an entry into another specified condition defined in the Applicability of a TS. In addition, pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. Thus, there will be no Actions that require the shutdown of a unit.

SR 3.0.5 addresses the application of SRs to the individual units, unless otherwise indicated. It is proposed for revision since unit-specific TS SRs will be applicable to the permanently defueled facility; however, the Technical Specification 3.7.8 SR has no exceptions.

Summary Page 50 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 3.0 of the PDTS, including LCO 3.0.1, LCO 3.0.2, SR 3.0.1, SR 3.0.2, SR 3.0.3, SR 3.0.4, and SR 3.0.5, 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. Therefore, retaining TS Section 3.0, as revised, provides appropriate control over use and application of the Dresden TS.

TS Section 3.1, Reactivity Control Systems TS Being Deleted TS Being Retained 3.1.1 - Shutdown Margin (SDM) 3.1.2 - Reactivity Anomalies 3.1.3 - Control Rod OPERABILITY 3.1.4 - Control Rod Scram Times 3.1.5 - Control Rod Scram Accumulators 3.1.6 - Rod Pattern Control 3.1.7 - Standby Liquid Control (SLC) System 3.1.8 - Scram Discharge Volume (SDV) Vent and Drain Valves TS Section 3.1 contains LCOs and SRs to assure and verify operability of reactivity control systems.

Once Exelon submits the certifications required by 10 CFR 50.82(a)(1) for Dresden, Units 2 and 3, 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). Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, reactivity control systems will not be required and these LCOs (and associated SRs) will not apply in a defueled condition. Therefore, TS Section 3.1 is proposed for deletion in its entirety.

Basis Technical Specification 3.1.1 defines the minimum shutdown margin for the reactor core in Modes 1, 2, 3, 4, and 5. SDM requirements are specified to ensure that, (a.) the reactor can be made subcritical from all operating conditions and transients and Design Basis Events; (b.) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and (c.) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. Technical Specification 3.1.1 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.2 reactivity anomaly limits are established to ensure plant operation is maintained within the assumptions of the safety analysis. Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the "Nuclear Design Methodology" are larger than expected.

Therefore, Reactivity Anomalies is used as a measure of the predicted versus measured core reactivity during power operation. This Specification is applicable during Modes 1 and 2. Technical Specification 3.1.2 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.3 This Specification defines the operability requirements for the control rods.

Control rods are components of the Control Rod Drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. This Specification is applicable during Modes 1 and 2. Technical Specification 3.1.3 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.4 defines the control rod scram times. The scram function of the CRD System controls reactivity changes during anticipated operational occurrences to ensure that acceptable Page 51 of 85

Attachment 1 Evaluation of Proposed Changes fuel design limits are not exceeded. The DBA and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits. Control rod scram times are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met. This Specification is applicable during Modes 1 and 2. Technical Specification 3.1.4 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.5 defines the operability requirements for the control rod scram accumulators.

The control rod scram accumulators are part of the CRD System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure. This Specification is applicable during Modes 1 and 2. Technical Specification 3.1.5 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.6 assures that the control rod patterns are consistent with the assumptions of the Control Rod Drop Accident analyses. Control rod patterns are controlled by the operator and the Rod Worth Minimizer so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% reactor thermal power (RTP). This Specification is applicable during Modes 1 and 2 with THERMAL POWER < 10% RTP. Technical Specification 3.1.6 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.7 defines the operability requirements for the SLC System. The SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The system is designed to bring the reactor from full power to a subcritical condition without taking credit for control rod movement. This Specification is applicable during Modes 1, 2, and 3. Technical Specification 3.1.7 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Technical Specification 3.1.8 defines the operability requirements for the SDV vent and drain valves. The SDV vent and drain valves discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. This Specification is applicable during Modes 1 and 2.

Technical Specification 3.1.8 does not apply once the reactors are permanently defueled; therefore, it is proposed to be deleted.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 5. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). Therefore, the TS listed in the previous paragraphs, which only address their associated specific plant equipment, control of process variables, design features, or operating restrictions are no longer applicable. Based on the above, the proposed deletion of all TS in Section 3.1, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

Page 52 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 3.2, Power Distribution Limits TS Being Deleted TS Being Retained 3.2.1 - Average Planar Linear Heat Generation Rate (APLHGR) 3.2.2 - Minimum Critical Power Ratio (MCPR) 3.2.3 - Linear Heat Generation Rate (LHGR)

TS Section 3.2 contains LCOs and SRs to ensure that power distribution limits are met. Once Exelon submits the certifications required by 10 CFR 50.82(a)(1) for Dresden, Units 2 and 3, 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). Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessels, power distribution limits will not be required and these LCOs (and associated SRs) will not apply in a defueled condition. Therefore, TS Section 3.2 is proposed for deletion in its entirety.

Basis Technical Specification 3.2.1 defines limits for the APLHGR. The APLHGR is a measure of the average Linear Heat Generation Rate of all the fuel rods in a fuel assemble at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during anticipated operational occurrences and that the peak cladding temperature during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. This specification is applicable when THERMAL POWER > 25% RTP.

Technical Specification 3.2.2 defines limits for the MCPR. MCPR is the ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The operating MCPR limit is established to ensure that no fuel damage results during anticipated operational occurrences. This Specification is applicable when THERMAL POWER > 25% RTP.

Technical Specification 3.2.3 defines limits for the LHGR. The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences and to ensure that the peak cladding temperature during the postulated design basis LOCA does not exceed the limits specified in 10 CFR 50.46. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. This specification is applicable when THERMAL POWER > 25% RTP.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when reactor thermal power is > 25% RTP. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 license 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). Therefore, the TS listed in the previous paragraphs, which only address their associated specific plant equipment, control of process variables, design features, or operating restrictions are no longer applicable. Based on the above, the proposed deletion of all TS in Section 3.2, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

Page 53 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 3.3, Instrumentation TS Being Deleted TS Being Retained 3.3.1.1 - Reactor Protection System (RPS) Instrumentation 3.3.1.2 - Source Range Monitor (SRM) Instrumentation 3.3.1.3 - Oscillation Power Range Monitor (OPRM)

Instrumentation 3.3.2.1 - Control Rod Block Instrumentation 3.3.2.2 - Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.3.1 - Post Accident Monitoring (PAM)

Instrumentation 3.3.4.1 - Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

Instrumentation 3.3.5.1 - Emergency Core Cooling System (ECCS)

Instrumentation 3.3.5.2 - Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation 3.3.5.3 - Isolation Condenser (IC) System Instrumentation 3.3.6.1 - Primary Containment Isolation Instrumentation 3.3.6.2 - Secondary Containment Isolation Instrumentation 3.3.6.3 - Relief Valve Instrumentation 3.3.7.1 - Control Room Emergency Ventilation (CREV)

System 3.3.7.2 - Mechanical Vacuum Pump Trip Instrumentation 3.3.8.1 - Loss of Power (LOP)

Instrumentation 3.3.8.2 - Reactor Protection System (RPS) Electric Power Monitoring TS Section 3.3 contains LCOs and SRs to assure and verify operability of instrumentation systems.

Once Exelon submits the certifications required by 10 CFR 50.82(a)(1) for Dresden, 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). Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessels, the Page 54 of 85

Attachment 1 Evaluation of Proposed Changes instrumentation addressed in TS Section 3.3 will not be required and these LCOs (and associated SRs) will not apply in a defueled condition. Therefore, TS Section 3.3 is proposed for deletion in its entirety.

Basis Technical Specification 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," provides the operability requirements for Instrumentation that ensures safe operation of the reactor by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. Technical Specification 3.3.1.1 is applicable in MODES 1, 2, and 5 as specified in TS Table 3.3.1.1-1.

Technical Specification 3.3.1.2, "Source Range Monitor (SRM) Instrumentation," provides operability requirements for the SRM instruments, which provide the operator with information relative to the neutron flux level at very low flux levels in the reactor core. This information is used by the operator to monitor the approach to criticality and determine when criticality is achieved. Technical Specification 3.3.1.2 is applicable in MODES 2, 3, 4, and 5 as specified in TS Table 3.3.1.2-1.

Technical Specification 3.3.1.3, "Oscillation Power Range Monitor (OPRM) Instrumentation," provides operability requirements for the OPRMs which detect and suppress neutron flux oscillations in the event of thermal-hydraulic instability. The region of anticipated oscillation is defined by THERMAL POWER

> 25% RTP and recirculation drive flow < 60% of rated recirculation drive flow. Therefore, this specification is applicable when THERMAL POWER is > 25% RTP.

Technical Specification 3.3.2.1, including TS Table 3.3.2.1-1, "Control Rod Block Instrumentation",

provides the operability requirements for the control rod block instrumentation specified in TS Table 3.3.2.1-1. The control rod block instrumentation is to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. Technical Specification 3.3.2.1 is applicable at various reactor thermal power levels and with the reactor: mode switch in the shutdown position as specified in TS Table 3.3.2.1-1.

Technical Specification 3.3.2.2, "Feedwater System and Main Turbine High Water Level Trip Instrumentation," provides operability requirements for the Feedwater System and Main Turbine High Water Level Trip Instrumentation that ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event.

This specification is applicable when THERMAL POWER is > 25% RTP.

Technical Specification 3.3.3.1, " Post Accident Monitoring (PAM) Instrumentation," provides the operability requirements for the PAM instrumentation specified in TS Table 3.3.3.1-1. The primary purpose of the PAM instrumentation is to display, in the control room, plant variables that provide information required by the control room operators during accident situations. This specification is applicable during Modes 1 and 2.

Technical Specification 3.3.4.1, "Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation," provides the operability requirements for instrumentation that initiates a Recirculation Pump trip to aid in preserving the integrity of the fuel cladding following events in which a scram does not, but should, occur. Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. This specification is applicable in Mode 1.

Technical Specification 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation," provides the operability requirements for Instrumentation that initiates the appropriate responses from emergency systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient.

For most Abnormal Operational Transients and Design Basis Accidents (DBAs), a wide range of Page 55 of 85

Attachment 1 Evaluation of Proposed Changes dependent and independent parameters are monitored. TS 3.3.5.1 is applicable in MODES 1, 2, and 3 in accordance with TS Table 3.3.5.1-1.

Technical Specification 3.3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation," supports the requirements of LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," and the definition of DRAIN TIME. The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCI). Technical Specification 3.3.5.2 is applicable in MODES 4 and 5 and when automatic isolation of an associated penetration flow path is credited in calculating DRAIN TIME in accordance with TS Table 3.3.5.2-1.

Technical Specification 3.3.5.3, "Isolation Condenser (IC) Instrumentation," provides the operability requirements for Instrumentation that initiates actions to provide adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser). Technical Specification 3.3.5.3 is applicable in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure > 150 psig.

Technical Specification 3.3.6.1, "Primary Containment Isolation Instrumentation," provides the operability requirements for Instrumentation that initiates automatic closure of appropriate Primary Containment Isolation Valves (PCIVs) to limit fission product release during and following postulated Design Basis Accidents. Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA. Technical Specification 3.3.6.1 is applicable in MODES 1, 2, and 3 in accordance with TS Table 3.3.6.1-1.

Technical Specification 3.3.6.2, "Secondary Containment Isolation Instrumentation," provides the operability requirements for Instrumentation that initiates automatic closure of appropriate Secondary Containment Isolation Valves (SCIVs) and starts the Standby Gas Treatment (SGT) System to limit fission product release during and following postulated DBAs. Secondary containment isolation and establishment of vacuum with the SGT System ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be operable are maintained within applicable limits. Technical Specification 3.3.6.2 is applicable in MODES 1, 2, and 3 in accordance with TS Table 3.3.6.2-1.

Technical Specification 3.3.6.3, "Relief Valve Instrumentation," ensures that the containment loads remain within the primary containment design basis. The opening setpoints of the relief valves also ensure that transient analyses assumptions can be met. Technical Specification 3.3.6.3 is applicable in MODES 1, 2, and 3.

Technical Specification 3.3.7.1, "Control Room Emergency Ventilation (CREV) System Instrumentation,"

provides the operability requirements for the Control Room Emergency Ventilation (CREV) System instrumentation, which ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits set by 10 CFR 50.67. This specification is applicable in Modes 1, 2, and 3, or during movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) in the secondary containment.

Technical Specification 3.3.7.2, "Mechanical Vacuum Pump Trip Instrumentation," provides the operability requirements for Instrumentation that initiates a trip of the main condenser mechanical vacuum pump breaker following events in which main steam line radiation exceeds predetermined values.

Tripping the mechanical vacuum pump limits the offsite and control room doses in the event of a control rod drop accident (CRDA). Technical Specification 3.3.7.2 is applicable in Modes 1 and 2 with the mechanical vacuum pump in service and any main steam line not isolated.

Page 56 of 85

Attachment 1 Evaluation of Proposed Changes Technical Specification 3.3.8.1, "Loss of Power (LOP) Instrumentation," provides the operability requirements for the LOP instrumentation specified in TS Table 3.3.8.1-1. This instrumentation monitors the 4160 V Essential Service System (ESS) buses. If the monitors determine that insufficient voltage is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources. Technical Specification 3.3.8.1 is applicable in Modes 1, 2, and 3, or when the associated diesel generator is required to be operable by LCO 3.8.2, "AC Sources -

Shutdown."

Technical Specification 3.3.8.2, "Reactor Protection System (RPS) Electric Power Monitoring," provides the operability requirements for the RPS electric power monitoring assemblies that isolate the RPS bus from the normal uninterruptible power supply (UPS) or alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This specification is applicable in Modes 1 and 2, or Mode 5 with any control rod withdrawn for a core cell containing one or more fuel assemblies.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 5, during movement of recently irradiated fuel in the secondary containment, or when DGs are required during Modes 4 and 5. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). Technical Specification 3.3.7.1 supports Technical Specification 3.7.4. As discussed below, Technical Specification 3.7.4 is proposed for deletion, therefore the instrumentation system is no longer needed to support it. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of decay following reactor shutdown, the nuclear fuel will no longer be considered "recently irradiated," and after 48 days of decay following permanent shutdown, the need for filtered recirculation of control room air will not be required, precluding the need for the DGs to support plant equipment associated with control room ventilation. Proposed License Conditions 2.C.(22) and 2.EE for Dresden, Unit 2 and Unit 3, respectively, will prohibit movement of irradiated fuel in the SFPs after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 48 days after permanent shutdown, effectively preventing the possibility of an FHA during this timeframe. Therefore, following permanent shutdown and defueling of the reactors, the Modes or conditions of applicability for these TS will no longer exist. Based on the above, the proposed deletion of all the TS in Section 3.3, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With these TS sections deleted in their entirety, their corresponding TS Bases will also be deleted.

TS Section 3.4, Reactor Coolant System (RCS)

TS Being Deleted TS Being Retained 3.4.1 - Recirculation Loops Operating 3.4.2 - Jet Pumps 3.4.3 - Safety and Relief Valves 3.4.4 - RCS Operational LEAKAGE 3.4.5 - RCS Leakage Detection Instrumentation 3.4.6 - RCS Specific Activity 3.4.7 - Shutdown Cooling (SDC) SystemHot Shutdown Page 57 of 85

Attachment 1 Evaluation of Proposed Changes 3.4.8 - Shutdown Cooling (SDC) SystemCold Shutdown 3.4.9 - RCS Pressure and Temperature (P/T) Limits 3.4.10 - Reactor Steam Dome Pressure TS Section 3.4 contains LCOs and SRs that provide assurance of the integrity and safe operation of the RCS and the operation of the related safety devices. Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, the LCOs (and associated SRs) will not apply (or are no longer needed) in a defueled condition. Therefore, TS Section 3.4 is proposed for deletion in its entirety.

Basis Technical Specification 3.4.1, "Recirculation Loops - Operating," provides the operability requirements for the Recirculation Loops. The Reactor Recirculation System provides forced coolant flow through the core to remove heat from the fuel. This Specification is applicable in Mode 1 and 2.

Technical Specification 3.4.2, "Jet Pumps," provides the operability requirements for the Jet pumps. The jet pumps are part of the Reactor Coolant Recirculation System and are designed to provide forced circulation through the core to remove heat from the fuel. This specification is applicable in Modes 1 and 2.

Technical Specification 3.4.3, "Safety and Relief Valves," provides the operability requirements for the Safety and Relief Valves. These valves provide overpressure protection to the reactor during operation.

This specification is applicable in Mode 1, 2, and 3.

Technical Specification 3.4.4, "RCS Operational LEAKAGE," provides the allowable leakage rates of reactor coolant from the RCS. The limits provided protection of the reactor coolant pressure boundary from degradation and the core from inadequate cooling. This specification is applicable in Mode 1, 2, and 3.

Technical Specification 3.4.5, "RCS Leakage Detection Instrumentation," provides the operability requirements for the RCS leakage detection instrumentation. This specification is applicable in Mode 1, 2, and 3.

Technical Specification 3.4.6, "RCS Specific Activity," provides limits regarding RCS specific activity.

This specification is applicable in Mode 1 or in Modes 2 and 3 with any main steam line not isolated.

Technical Specification 3.4.7, "Shutdown Cooling (SDC) System - Hot Shutdown," provides the operability requirements for the SDC system during hot shut down. This specification is applicable in Mode 3 with reactor steam dome pressure less than the SDC cut-in permissive temperature.

Technical Specification 3.4.8, "Shutdown Cooling (SDC) System - Cold Shutdown," provides the operability requirements for the SDC system during cold shut down. This specification is applicable in Mode 4.

Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits," provides RCS P/T limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary. This specification limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. This specification is applicable at all times.

Page 58 of 85

Attachment 1 Evaluation of Proposed Changes Technical Specification 3.4.10, "Reactor Steam Dome Pressure," provides the limit on the reactor steam dome pressure. The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of DBAs and transients. This specification is applicable in Mode 1 and 2.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 4. Technical Specification 3.4.9 applies to pressure and temperature limits of the reactor regardless of MODE. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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).

Therefore, the TS listed in the previous paragraphs, which only address their associated specific plant equipment, control of process variables, design features, or operating restrictions are no longer applicable. With the reactors in a permanently shutdown and defueled condition, reactor pressure and temperature limits do not apply. Based on the above, the proposed deletion of all TS in Section 3.4, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility.

With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

TS Section 3.5, Emergency Core Cooling Systems (ECCS), Reactor Pressure Vessel (RPV) Water Inventory Control, and Isolation Condenser (IC) System TS Being Deleted TS Being Retained 3.5.1 - ECCSOperating 3.5.2 - RPV Water Inventory Control 3.5.3 - Isolation Condenser (IC) System TS Section 3.5 contains LCOs that provide for appropriate control of process variables, design features, or operating restrictions needed for appropriate functional capability of RCS equipment required for safe operation of the facility.

The TS listed below do not apply once the reactor is permanently defueled; therefore, their corresponding LCOs, and associated SRs, are proposed for deletion. Similarly, the corresponding TS Bases are also proposed for deletion.

Basis Technical Specification 3.5.1, "ECCS - Operating," provides the operability requirements for the ECCS, which is designed to limit the release of radioactive materials to the environment following a Loss of Coolant Accident (LOCA). The ECCS network consists of the High Pressure Coolant Injection (HPCI)

System, the Core Spray (CS) System, the Low Pressure Coolant Injection (LPCI) System, and the ADS.

This specification is applicable in MODE 1 and in MODES 2 and 3, except High Pressure Coolant Injection (HPCI) and Automatic Depressurization System (ADS) valves are not required to be operable with the reactor steam dome pressure 150 psig.

Technical Specification 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," provides the operability requirements that ensure the RPV water level remains above the top of the active irradiated fuel at all times to prevent elevated fuel cladding temperatures when the reactor is in cold shutdown or refueling. Technical Specification 3.5.2 is applicable in MODES 4 and 5.

Technical Specification 3.5.3, "IC System," provides operability requirements for the IC system, which is designed to respond to transient events by providing adequate core cooling following RPV isolation. This specification is applicable in Mode 1 or Modes 2 and 3 with reactor steam dome pressure > 150 psig.

Page 59 of 85

Attachment 1 Evaluation of Proposed Changes Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 5. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). Therefore, the TS listed in the previous paragraphs, which only address their associated specific plant equipment, control of process variables, design features, or operating restrictions are no longer applicable. Based on the above, the proposed deletion of all TS in Section 3.5, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

TS Section 3.6, Containment Systems TS Being Deleted TS Being Retained 3.6.1.1 - Primary Containment 3.6.1.2 - Primary Containment Air Lock 3.6.1.3 - Primary Containment Isolation Valves (PCIVs) 3.6.1.4 - Drywell Pressure 3.6.1.5 - Drywell Air Temperature 3.6.1.6 - Low Set Relief Valves 3.6.1.7 - Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.8 - Suppression Chamber-to-Drywell Vacuum Breakers 3.6.2.1 - Suppression Pool Average Temperature 3.6.2.2 - Suppression Pool Water Level 3.6.2.3 - Suppression Pool Cooling 3.6.2.4 - Suppression Pool Spray 3.6.2.5 - Drywell-to-Suppression Chamber Differential Pressure 3.6.3.1 - Primary Containment Oxygen Concentration 3.6.4.1 - Secondary Containment 3.6.4.2 - Secondary Containment Isolation Valves (SCIVs) 3.6.4.3 - Standby Gas Treatment (SGT) System Page 60 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 3.6 contains LCOs and SRs that provide assurance of the integrity and safe operation of the containment systems. Following the submittal of certifications of permanently defueled conditions in accordance with 10 CFR 50.82(a)(1), the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessels, the LCOs (and associated SRs) will not apply (or are no longer needed) in a defueled condition. Therefore, TS Section 3.6 is proposed for deletion in its entirety.

Basis Technical Specification 3.6.1.1, "Primary Containment," provides operability requirements for the primary containment. Its function was to isolate and contain fission products released from the Reactor Primary System following a DBA and to confine the postulated release of radioactive material to within limits. This specification is applicable in Modes 1, 2 and 3.

Technical Specification 3.6.1.2, "Primary Containment Air Lock," provides operability requirements for the primary containment air lock. The airlock provides personnel access to the primary containment and provides containment isolation during the process of personnel entry and exit. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)," provides operability requirements for the PCIVs. Their function was to limit fission products released during and following a DBA to within limits. This specification is applicable in Modes 1, 2, and 3 when associated instrumentation is required to be OPERABLE.

Technical Specification 3.6.1.4, "Drywell Pressure," provides a limit regarding Drywell pressure during normal operation to preserve the initial conditions assumed in the accident analysis for a DBA. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.1.5, "Drywell Air Temperature," provides a limit regarding Drywell air temperature to preserve the initial conditions assumed in the accident analysis for a DBA. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.1.6, "Low Set Relief Valves," provides operability requirements for the low set relief function of two of the relief valves. The low set relief function prevents excessive short duration relief cycles with valve actuation at the low set relief setpoint. The requirements of this LCO are applicable to the two Electromatic relief valves that contain the low set relief function for controlling the opening and closing of the relief valves. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.1.7, "Reactor Building-to-Suppression Chamber Vacuum Breakers," provides operability requirements for the Reactor Building-to-Suppression Chamber Vacuum Breakers. Its function was to relieve vacuum when primary containment depressurizes below reactor building pressure.

The reactor building-to-suppression chamber vacuum breakers prevent an excessive negative differential pressure across the primary containment boundary. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.1.8, "Suppression Chamber-to-Drywell Vacuum Breakers," specification provides operability requirements for the Suppression Chamber-to-Drywell Vacuum Breakers. Its function was to relieve vacuum in the Drywell. The suppression chamber-to-Drywell vacuum breakers prevent an excessive negative differential pressure across the suppression chamber-drywell boundary.

This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.2.1, "Suppression Pool Average Temperature," provides a limit regarding suppression pool average temperature to assure that the primary containment conditions assumed for the safety analysis are met. This specification is applicable in Modes 1, 2, and 3.

Page 61 of 85

Attachment 1 Evaluation of Proposed Changes Technical Specification 3.6.2.2, "Suppression Pool Water Level," provides a limit regarding suppression pool water level to preserve the initial conditions assumed in the accident analysis for a DBA. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.2.3, "Suppression Pool Cooling," provides operability requirements for the suppression pool cooling system. Its function was to remove heat from the suppression pool following a DBA. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.2.4, "Suppression Pool Spray," provides operability requirements for the suppression pool spray system. The function of the suppression pool spray system is to absorb the sudden input of heat from the primary system in the suppression chamber air space from a DBA or a rapid depressurization of the reactor pressure vessel (RPV) through relief valves. This specification is applicable in Modes 1, 2, and 3.

Technical Specification 3.6.2.5, "Drywell-to-Suppression Chamber Differential Pressure," provides requirements for the Drywell-to-Suppression Chamber Differential Pressure. The purpose of maintaining the drywell at a slightly higher pressure with respect to the suppression chamber is to minimize the drywell pressure increase necessary to clear the downcomer pipes to commence condensation of steam in the suppression pool and to minimize the mass of the accelerated water leg. This reduces the hydrodynamic loads on the torus downcomer piping during the LOCA blowdown. This specification is applicable in Mode 1 during the time period from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Technical Specification 3.6.3.1, "Primary Containment Oxygen Concentration," provides requirements for the primary containment oxygen concentration. The function of the primary containment oxygen concentration is to ensure that an event that produces any amount of hydrogen and oxygen does not result in a combustible mixture inside primary containment. This specification is applicable in Mode 1 during the time period from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Technical Specification 3.6.4.1, "Secondary Containment," provides the operability requirements for secondary containment. The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a DBA. In conjunction with operation of the SGT System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be operable, or that take place outside primary containment. This specification is applicable in Modes 1, 2, and 3, or during movement of recently irradiated fuel assemblies in the secondary containment.

Technical Specification 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," provides the operability requirements for the SCIVs. The function of the SCIVs is to limit fission product release during and following postulated DBAs. This specification is applicable in Modes 1, 2, and 3, or during movement of recently irradiated fuel in the secondary containment.

Technical Specification 3.6.4.3," Standby Gas Treatment (SGT) System," provides the operability requirements for the SGT system. The function of the SGT system is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a DBA are filtered and adsorbed in charcoal filters prior to exhausting to the environment. This specification is applicable in Modes 1, 2, and 3, or during movement of recently irradiated fuel in the secondary containment.

Page 62 of 85

Attachment 1 Evaluation of Proposed Changes Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 3 or during movement of recently irradiated fuel in the secondary containment. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). As discussed in the Fuel Handling Accident Analysis for the Permanently Defueled Condition section of this attachment, in the post permanent shutdown FHA analysis 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 with the unit permanently defueled. Therefore, the use of these systems is not credited or required in the FHA for reduction of nuclides or a reduction of onsite or offsite doses after 48 days of decay time. Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessels until 48 days after permanent shutdown through the imposition of proposed License Conditions 2.C.(22) and 2.EE for Dresden Unit 2 and 3, respectively. This will effectively prevent an FHA from occurring until after the 48-day decay period has elapsed and allows these LCOs to be eliminated during the decay period. Additionally, after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of decay following reactor shutdown, the nuclear fuel will no longer be considered "recently irradiated;" therefore, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, the Modes or conditions of applicability for these TS will no longer exist. Based on the above, the proposed deletion of all the TS in Section 3.6, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With these TS sections deleted in their entirety, their corresponding TS Bases will also be deleted.

TS SECTION 3.7, Plant Systems (Current Title)

TS Section 3.7, Facility Systems (Proposed Title)

TS Being Deleted TS Being Retained 3.7.1 - Containment Cooling Service Water (CCSW)

System 3.7.2 - Diesel Generator Cooling Water (DGCW)

System 3.7.3 - Ultimate Heat Sink (UHS) 3.7.4 - Control Room Emergency Ventilation (CREV) System 3.7.5 - Control Room Emergency Ventilation Air Conditioning (AC) System 3.7.6 - Main Condenser Offgas 3.7.7 - Main Turbine Bypass System 3.7.8 - Spent Fuel Storage Pool Water Level TS Section 3.7 contains LCOs and SRs that provide assurance of the safe operation of various plant systems. Because, following the submittal of certifications of permanently defueled conditions in accordance with 10 CFR 50.82(a)(1), the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, the LCOs (and associated SRs) that do not apply (or are no longer needed) in a defueled condition are being proposed for deletion.

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.

Technical Specification 3.7.8, "Spent Fuel Storage Pool Water Level," is proposed for retention in the PDTS with the changes described below.

Basis Page 63 of 85

Attachment 1 Evaluation of Proposed Changes Technical Specification 3.7.1, "Containment Cooling Service Water (CCSW) System," provides the operability requirements for the CCSW System. The CCSW System is to provide cooling water for the containment cooling heat exchangers, required for a safe reactor shutdown following a DBA or transient.

The CCSW System is operated whenever the containment cooling heat exchangers are required to operate in the suppression pool cooling mode or in the containment spray mode of the LPCI System.

This specification is applicable during Modes 1, 2 and 3.

Technical Specification 3.7.2, "Diesel Generator Cooling Water (DGCW) System," provides the operability requirements for the DGCW System. The function of the DGCW System is to provide cooling water for the removal of heat from the two diesel generator (DG) heat exchangers. The DGCW system can also be used as an alternate water supply for CCSW keep fill. This specification is applicable during Modes 1, 2 and 3.

Technical Specification 3.7.3, "Ultimate Heat Sink (UHS)," provides the operability requirements for the UHS. The function of the UHS is to provide a suction source and discharge pathway for the cooling water associated with the CCSW and DGCW Systems. The UHS consists of water sources from either the Kankakee River (normal), or the cooling lake (alternate) and can be aligned as either a closed cycle operating system utilizing the cooling lake and canals, or an open cycle operating system with the discharge returning to the Illinois River. This specification is applicable during Modes 1, 2 and 3.

Technical Specification 3.7.4, "Control Room Emergency Ventilation (CREV) System," provides the operability requirements for the CREV System. The function of the CREV System is to provide a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. This specification is applicable during Modes 1, 2 and 3, or during movement of recently irradiated fuel assemblies in the secondary containment.

Technical Specification 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System,"

provides the operability requirements for the Control Room Emergency Ventilation AC system. The function of the Control Room Emergency Ventilation AC system is to provide temperature control for the control room emergency zone following isolation of the control room emergency zone. The Control Room Emergency Ventilation AC System is designed to provide a controlled environment under both normal and accident conditions. This specification is applicable during Modes 1, 2 and 3, or during movement of recently irradiated fuel assemblies in the secondary containment. Proposed License Conditions 2.C.(22) and 2.EE for Dresden, Unit 2 and Unit 3, respectively, will prohibit movement of irradiated fuel in the SFPs after the submittal of the certifications of permanent removal of fuel from the reactor vessels until 48 days after permanent shutdown, effectively preventing the possibility of an FHA during this timeframe.

Technical Specification 3.7.6, "Main Condenser Offgas," provides the operability requirements for the Main Condenser Offgas System. The function of the Main Condenser Offgas System is to reduce the gaseous radwaste emission. This specification is applicable during Mode 1 or in Modes 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.

Technical Specification 3.7.7, "The Main Turbine Bypass System," provides the operability requirements for the Main Turbine Bypass System. The function of the Main Turbine Bypass System is to control steam pressure when reactor steam generation exceeds turbine requirements during plant startup, sudden load reduction, and cool down. This specification is applicable when THERMAL POWER is

> 25% RTP.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 3 or during movement of recently irradiated fuel in the Page 64 of 85

Attachment 1 Evaluation of Proposed Changes secondary containment. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). As discussed in the Fuel Handling Accident Analysis for the Permanently Defueled Condition section of this attachment, in the post permanent shutdown FHA analysis 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 with the units permanently shutdown and defueled. Therefore, the use of these systems is not credited or required in the FHA for reduction of nuclides or a reduction of onsite or offsite doses after 48 days of decay time. Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 days after permanent shutdown through the imposition of the proposed License Conditions 2.C.(22) and 2.EE for Dresden, Units 2 and 3, respectively. This will effectively prevent an FHA from occurring until after the 48-day decay period has elapsed and allows these LCOs to be eliminated following permanent defueling during the decay period. Additionally, after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of decay following reactor shutdown, the nuclear fuel will no longer be considered "recently irradiated;" therefore, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, the Modes or conditions of applicability for these TS will no longer exist. Based on the above, the proposed deletion of all these TS, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With these TS sections deleted in their entirety, their corresponding TS Bases will also be deleted.

Technical Specification 3.7.8, "Spent Fuel Storage Pool Water Level," provides the minimum water level in the SFPs and applies whenever movement of irradiated fuel assemblies occurs in the spent fuel storage pool and during movement of new fuel assemblies in the spent fuel storage pool with irradiated fuel assemblies seated in the spent fuel storage pool. The water level above the irradiated fuel assemblies is an explicit assumption of the fuel handling accident. A fuel handling accident is evaluated to ensure that the radiological consequences (calculated control room operator dose and doses at the exclusion area and low population zone boundaries) are below the 10 CFR 50.67 exposure guidelines.

This specification will be retained in the PDTS with the following proposed changes. The applicability related to the movement of new fuel assemblies in the spent fuel pool is proposed for deletion, as no new fuel will be moved for the permanently shutdown and defueled condition. The Note in REQUIRED ACTION A.1 is deleted, because it states that LCO 3.0.3 is not applicable; however, LCO 3.0.3 will no longer exist in the PDTS as discussed in the changes proposed to TS Section 3.0. Lastly, the frequency for SR 3.7.8.1 is proposed to be modified to account for the proposed elimination of the Surveillance Frequency Control Program. The proposed changes to this specification are shown below and in . Proposed changes to the TS Bases for this specification are shown in Attachment 4 for information only.

Page 65 of 85

Attachment 1 Evaluation of Proposed Changes Proposed Changes to Technical Specification 3.7.8 Page 66 of 85

Attachment 1 Evaluation of Proposed Changes 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 and Starting Air 3.8.4 - DC Sources-Operating 3.8.5 - DC Sources-Shutdown 3.8.6 - Battery Parameters 3.8.7 - Distribution Systems Operating 3.8.8 - Distribution Systems Shutdown TS Section 3.8 contains LCOs and SRs that provide assurance of the integrity and safe operation of AC Sources, DC Sources, and Electrical Distribution Systems. Because, following the submittal of certifications of permanently defueled conditions in accordance with 10 CFR 50.82(a)(1), the Dresden, Units 2 and 3 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, the LCOs (and associated SRs) will not apply (or are no longer needed) in a defueled condition.

Therefore, TS Section 3.8 is proposed for deletion in its entirety.

Basis Technical Specification 3.8.1, "AC Sources - Operating," provides the operability requirements for AC sources during specific operating Modes (i.e., Modes 1, 2, and 3). The function of the AC electrical power sources is to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to engineered safeguards systems so that the fuel, RCS, and containment design limits are not exceeded.

Technical Specification 3.8.2, "AC Sources - Shutdown," provides the operability requirements for AC sources during specific shutdown Modes (i.e., Modes 4, 5 or during movement of recently irradiated fuel assemblies in the secondary containment). The function of the AC electrical power sources is to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to engineered safeguards systems so that the fuel, RCS, and containment design limits are not exceeded.

Technical Specification 3.8.3, "Diesel Fuel Oil and Starting Air," provides limits on the Diesel Fuel Oil and Starting Air Subsystems. These systems support the operation of the standby AC power sources in accordance with TS 3.8.1 and 3.8.2. This specification is applicable when the associated Diesel Generators are required to be operable.

Technical Specification 3.8.4, "DC Sources - Operating," provides the operability requirements for DC sources during specific operating Modes (i.e., Modes 1, 2, and 3). The function of the DC electrical power sources is to provide the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment.

Technical Specification 3.8.5, "DC Sources - Shutdown," provides the operability requirements for DC sources during specific shutdown Modes (i.e., Modes 4, 5 and during movement of recently irradiated fuel assemblies in the secondary containment). The function of the DC electrical power sources is to provide the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment.

Technical Specification 3.8.6, "Battery Parameters," provides limits on various battery parameters (i.e.,

battery float current, electrolyte temperature, level and float voltage). These support the batteries that are Page 67 of 85

Attachment 1 Evaluation of Proposed Changes required to be operable in accordance with Technical Specification 3.8.4 and Technical Specification 3.8.5.

Technical Specification 3.8.7, "Distribution Systems - Operating," provides the operability requirements for AC and DC distribution systems during specific operating Modes (i.e., Modes 1, 2, and 3). The AC and DC 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, Reactor Coolant System, and containment design limits are not exceeded.

Technical Specification 3.8.8, "Distribution Systems - Shutdown," provides the operability requirements for AC and DC distribution systems during specific shutdown Modes (i.e., Modes 4 and 5) or during movement of recently irradiated fuel assemblies in the secondary containment). The function of the AC and DC and uninterruptible AC bus electrical power distribution systems provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to Engineered Safety Feature systems so that the fuel, RCS, and containment design limits are not exceeded.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility only when the reactor is in MODES 1 through 5, during movement of recently irradiated fuel in the secondary containment, or when DGs are required to be operable. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2).

As discussed in the Fuel Handling Accident Analysis for the Permanently Defueled Condition section of this attachment, in the post permanent shutdown FHA analysis 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 with the units permanently shutdown and defueled. Therefore, the use of these systems is not credited or required in the FHA for reduction of nuclides or a reduction of onsite or offsite doses after 48 days of decay time. Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 days after permanent shutdown through the imposition of the proposed License Conditions 2.C.(22) and 2.EE for Dresden, Units 2 and 3, respectively. This will effectively prevent an FHA from occurring until after the 48-day decay period has elapsed and allows these LCOs to be eliminated following permanent defueling during the decay period.

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of decay following reactor shutdown, the nuclear fuel will no longer be considered "recently irradiated," and after 48 days of decay following permanent shutdown, the need for filtered recirculation of control room air will not be required, precluding the need for the DGs to support plant equipment associated with control room ventilation.

Therefore, following permanent shutdown and defueling, the Modes or conditions of applicability for these TS will no longer exist. Based on the above, the proposed deletion of all the TS in Section 3.8, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With these TS sections deleted in their entirety, their corresponding TS Bases will also be deleted.

TS Section 3.9, Refueling Operations TS Being Deleted TS Being Retained 3.9.1 - Refueling Equipment Interlocks 3.9.2 - Refuel Position One-Rod-Out Interlock 3.9.3 - Control Rod Position Page 68 of 85

Attachment 1 Evaluation of Proposed Changes 3.9.4 - Control Rod Position Indication 3.9.5 - Control Rod OPERABILITYRefueling 3.9.6 - Reactor Pressure Vessel (RPV) Water Level Irradiated Fuel 3.9.7 - Reactor Pressure Vessel (RPV) Water Level New Fuel or Control Rods 3.9.8 - Shutdown Cooling (SDC)High Water Level 3.9.9 - Shutdown Cooling (SDC)Low Water Level TS Section 3.9 contains LCOs and SRs that provide operability requirements for systems and components and limits to: 1) prevent reactivity excursions; 2) limit offsite doses from an accident; and 3) remove decay heat from the RCS. These specifications are applicable during refueling operations.

Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, the LCOs (and associated SRs) in Section 3.9 will no longer apply in a defueled condition. Therefore, TS Section 3.9 is proposed for deletion in its entirety.

Basis Technical Specification 3.9.1, "Refueling Equipment Interlocks," provides the operability requirements for the refueling equipment interlocks. This specification is applicable during in-vessel fuel movement with the reactor mode switch in the refuel position. The function of these interlocks is to restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce plant procedures that prevent the reactor from achieving criticality during refueling.

Technical Specification 3.9.2, "Refuel Position One-Rod-Out Interlock," provides the operability requirements for the refuel position one-rod-out interlock. Its function was to restrict the movement of control rods to reinforce plant procedures that prevent the reactor from becoming critical during refueling operations (i.e., Mode 5).

Technical Specification 3.9.3, "Control Rod Position," requires all control rods to be fully inserted when loading fuel assemblies into the core to minimize the probability of an inadvertent criticality during refueling. This specification is applicable when loading fuel assemblies into the core.

Technical Specification 3.9.4, "Control Rod Position Indication," provides operability requirements for the control rod full-in position indication channel for each control rod to provide the required inputs to the refueling interlocks (i.e., TS 3.9.1 and 3.9.2) to prevent inadvertent criticalities during refueling operations.

This specification is applicable in Mode 5.

Technical Specification 3.9.5, "Control Rod OPERABILITY - Refueling," in conjunction with the Reactor Protection System, the Control Rod Drive (CRD) System provides the means for the reliable control of reactivity changes during refueling operation. In addition, the control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. This specification is applicable in Mode 5.

Technical Specification 3.9.6, "Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel," provides a limit regarding the RPV water level, and is applicable during movement of irradiated fuel assemblies within the RPV.

Page 69 of 85

Attachment 1 Evaluation of Proposed Changes Technical Specification 3.9.7, "Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods,"

provides a limit regarding the RPV water level and is applicable during movement of new fuel assemblies or handling of control rods within the RPV when fuel assemblies seated within the RPV are irradiated.

Technical Specification 3.9.8, "Shutdown Cooling (SDC) - High Water Level," provides operability requirements for the SDC System in Mode 5, which ensures decay and sensible heat is removed from the reactor coolant. This specification is applicable during Mode 5 with irradiated fuel in the RPV and the water level 23 ft. above the RPV flange.

Technical Specification 3.9.9, "Shutdown Cooling (SDC) - Low Water Level," provides operability requirements for the SDC System during Mode 5 with irradiated fuel in the RPV and the water level

< 23 ft above the RPV flange. The purpose of the RHR System in Mode 5 is to remove decay heat and sensible heat from the reactor coolant.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe refueling operation of the facility only when the reactor is in MODE 5 or during movement of fuel assemblies within the RPV. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, 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). Consequently, entry into MODE 5 and fuel movement within the RPV are not authorized. Therefore, the TS listed in the previous paragraphs, which only address these specific plant systems, control of process variables, design features, or operating restrictions are no longer applicable. Based on the above, the proposed deletion of all TS in Section 3.9, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

TS Section 3.10, Special Operations TS Being Deleted TS Being Retained 3.10.1 - Reactor Mode Switch Interlock Testing 3.10.2 - Single Control Rod Withdrawal - Hot Shutdown 3.10.3 - Single Control Rod Withdrawal - Cold Shutdown 3.10.4 - Single Control Rod Drive (CRD) Removal -

Refueling 3.10.5 - Multiple Control Rod Withdrawal - Refueling 3.10.6 - Control Rod Testing -

Operating 3.10.7 - SHUTDOWN MARGIN (SDM) Test - Refueling 3.10.8 - Inservice Leak and Hydrostatic Testing Operation TS Section 3.10 contains Special Operations LCOs and SRs that provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. Because the Dresden, Units 2 and 3, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessel, the Page 70 of 85

Attachment 1 Evaluation of Proposed Changes LCOs (and associated SRs) will no longer apply in a defueled condition. Therefore, TS Section 3.10 is proposed for deletion in its entirety.

Basis Technical Specification 3.10.1, "Reactor Mode Switch Interlock Testing," provides the requirements to permit operation of the reactor mode switch from one position to another to confirm certain aspects of associated interlocks during periodic tests and calibrations. This specification is applicable in Modes 3, 4, and 5.

Technical Specification 3.10.2, "Single Control Rod Withdrawal - Hot Shutdown," provides the requirements and additional controls to permit the withdrawal of a single control rod for testing while in hot shutdown by imposing certain restrictions. This specification is applicable in Mode 3 with the reactor mode switch in the refuel position.

Technical Specification 3.10.3, "Single Control Rod Withdrawal - Cold Shutdown," provides the requirements to permit the withdrawal of a single control rod for testing or maintenance, while in cold shutdown by imposing certain restrictions. This specification is applicable in Mode 4 with the reactor mode switch in the Refuel position.

Technical Specification 3.10.4, "Single Control Rod Drive (CRD) Removal - Refueling," provides the requirements to permit the removal of a single CRD during refueling operations by imposing certain administrative controls. This specification is applicable in Mode 5 with LCO 3.9.5 not met.

TS 3.10.5, "Multiple Control Rod Withdrawal - Refueling," provides the requirements to permit multiple control rod withdrawal during refueling by imposing certain administrative controls. This TS is Applicable in Mode 5 with TS LCOs 3.9.3, 3.9.4, or 3.9.5 not met.

Technical Specification 3.10.6, "Control Rod Testing - Operating," provides the requirements to permit control rod testing, while in Modes 1 and 2, by imposing certain administrative controls. This Special Operations LCO provides the necessary exceptions to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6. This specification is applicable in Modes 1 and 2 with LCO 3.1.6 not met.

Technical Specification 3.10.7, "SHUTDOWN MARGIN (SDM) Test - Refueling," provides the requirements to permit SDM testing to be performed for those plant configurations in which the RPV head is either not in place or the head bolts are not fully tensioned. This specification is applicable in Mode 5 with the reactor mode switch in the startup/hot standby position.

Technical Specification 3.10.8, "Inservice Leak and Hydrostatic Testing Operation," provides the requirements to allow certain reactor coolant pressure tests to be performed with the RPV at temperatures > 212°F (normally corresponding to Mode 3) or to allow completing these reactor coolant pressure tests when the initial conditions do not require temperatures > 212°F. Furthermore, the purpose is to allow continued performance of control rod scram time testing required by SR 3.1.4.1 or SR 3.1.4.4 if reactor coolant temperatures exceed 212°F when the control rod scram time testing is initiated in conjunction with an inservice leak or hydrostatic test. This specification is applicable in Mode 4 with average reactor coolant temperature > 212°F.

Summary The above TS are related to assuring the appropriate functional capability of plant equipment, and control of process variables, design features, or operating restrictions required for safe operation of the facility during specific operating scenarios. After the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the reactors or emplacement or Page 71 of 85

Attachment 1 Evaluation of Proposed Changes retention of fuel in the reactor vessels pursuant to 10 CFR 50.82(a)(2). Therefore, the TS listed in the previous paragraphs, which only address their associated specific plant equipment, control of process variables, design features, or operating restrictions are no longer applicable. Based on the above, the proposed deletion of all TS in Section 3.10, including associated SRs, is acceptable with no impact on continued safe maintenance of the facility. With the TS section deleted in its entirety, the corresponding TS Bases will also be deleted.

TS Section 4.0, Design Features TS Being Deleted TS Being Retained 4.1 - Site Location 4.2 - Reactor Core 4.3 - Fuel Storage The existing TS Section 4.0, "Design Features," contains descriptions and requirements for those features of the facility such as materials of construction and geometric arrangements which, if altered or modified, could have a significant effect on safety and are not covered in the previous sections of the TS.

Because the Dresden, 10 CFR 50 licenses will no longer authorize emplacement or retention of fuel in the reactor vessels, the design features that do not apply in a defueled condition are being proposed for deletion.

Current TS Proposed TS 4.1.1 Site and Exclusion Area Boundaries 4.1.1 Site and Exclusion Area Boundaryies The site area boundary follows the Illinois The site area boundary follows the Illinois River to the north, the Kankakee River to the River to the north, the Kankakee River to the east, a country road from Divine extended east, a country road from Divine extended eastward to the Kankakee River on the eastward to the Kankakee River on the south, and the Elgin, Joliet, and Eastern south, and the Elgin, Joliet, and Eastern Railway right-of-way on the west. The Railway right-of-way on the west. The exclusion area boundary shall be an exclusion area boundary shall be an 800 meter radius from the centerline of the 800 meter radius from the centerline of the reactor vessels. reactor vessels.

4.1.2 Low Population Zone 4.1.2 Low Population ZoneDeleted The low population zone shall be a five mile The low population zone shall be a five mile radius from the centerline of the reactor radius from the centerline of the reactor vessels. vessels.

Basis Technical Specification 4.1, "Site Location," describes the Dresden Site, Exclusion Area, and Low Population Zone Boundaries. In accordance to 10 CFR 50.82(a)(2), the facility licenses for Dresden 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 Exclusion Area and Low Population Zone relative to the reactor vessel is inconsequential. Therefore, TS 4.1 will be revised to remove the reference to the Exclusion Area Boundary and the Low Population Zone.

Technical Specification 4.2, "Reactor Core," provides descriptions of and requirements for fuel assemblies and control rod assemblies in the Dresden, reactor cores during operation of the units. In accordance with 10 CFR 50.82(a)(2), the facility licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel in the reactors. With the reactors permanently shutdown Page 72 of 85

Attachment 1 Evaluation of Proposed Changes and defueled, the operational reactor cores, as described in TS 4.2, will no longer exist. Therefore, this specification will be deleted in its entirety. Because the TS will not be renumbered, a markup is provided in Attachment 3 to identify this section as deleted.

Technical Specification 4.3, "Fuel Storage," provides a description and requirements for prevention of criticality, drainage, and capacity pertaining to irradiated fuel storage for Dresden, Units 2 and 3, SFPs.

The requirements for prevention of criticality of irradiated fuel, prevention of spent fuel pool drainage, and irradiated fuel capacity limitations will remain unchanged in the PDTS.

Summary Technical Specification 4.1 is being retained and revised. The information contained in Technical Specification 4.1.2 is inconsequential with the reactors permanently shutdown and defueled; therefore, this Technical Specification 4.1.2 is proposed for deletion.

Technical Specification 4.2 is proposed for deletion is in its entirety, as discussed above.

Technical Specification 4.3 will remain applicable with the reactors permanently defueled. As such, this TS section is being retained to reflect a permanently defueled condition.

TS Section 5.0, Administrative Controls TS Being Deleted TS Being Retained 5.1 - Responsibility 5.2 - Organization 5.3 - Unit Staff Qualifications 5.4 - Procedures 5.5 - Programs and Manuals 5.6 - Reporting Requirements 5.7 - High 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 Section 5.1 - Exelon proposed changes to the administrative controls in TS Section Responsibility 5.1 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3).

TS Section 5.2 - Exelon proposed changes to the administrative controls in TS Section Organization 5.2 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3).

TS Section 5.3 - Unit Staff Exelon proposed changes to the administrative controls in TS Section Qualifications 5.3 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3).

TS Section 5.4 - Procedures Exelon proposed changes to the administrative controls in TS Section 5.4 to reflect the permanently shutdown and defueled state in a letter dated September 24, 2020 (Reference 3).

TS Section 5.5.2 - Primary This program was established to provide controls to minimize leakage Coolant Sources Outside from those portions of systems outside containment that could contain Containment highly radioactive fluids during a serious transient or accident to levels as low as practicable. It addresses the Core Spray, HPCI, LPCI, IC, SDC, RWCU, Process Sampling, Containment Monitoring, and SGT Page 73 of 85

Attachment 1 Evaluation of Proposed Changes Systems. As previously discussed, the TS requirements for these systems were proposed for deletion. Once Dresden, Units 2 and 3, are permanently shutdown and defueled, there will no longer be any transient or accident conditions associated with primary coolant sources. Therefore, Technical Specification 5.5.2does not apply in the permanently shutdown and defueled condition, and will not be retained.

TS Section 5.5.4 - This program conforms to 10 CFR 50.36a for the control of radioactive Radioactive Effluent Controls effluents and for maintaining the doses to members of the public from Program 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 Section 5.5.5 - This program provides controls to track the cyclic and transient Component Cyclic or occurrences to ensure that components are maintained within the Transient Limit design limits. The program will not be retained in the PDTS, because the Dresden, 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 vessel will no longer be subjected to cycles or transients after permanent shutdown.

TS Section 5.5.7 - Ventilation This program was established to implement the required testing of the Filter Testing Program (VFTP) Engineered Safety Feature (ESF) filter ventilation systems (i.e., SGT and CREV 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 3.6.4.3 and 3.7.4 that provided the operability requirements for the SGT and CREV Systems are proposed to be eliminated.

TS Section 5.5.8 - Explosive This program provides controls for potentially explosive gas mixtures Gas and Storage Tank contained in the Radioactive Waste Disposal System, and the quantity Radioactivity Monitoring of radioactivity contained in gas storage tanks or fed into the Offgas Program Treatment System.

The requirements regarding the explosive gas portion of this program will not be retained. The requirements regarding explosive gas mixtures contained in the Main Condenser Offgas System are not applicable following permanent shutdown and defueling and are proposed for deletion from this specification. Technical Specification 3.7.6 provides the operability requirements for the Main Condenser Offgas System and as described above; it is not proposed to be included in the PDTS. Since the Main Condenser Offgas System will not be required once Dresden, Units 2 and 3, are permanently shutdown and defueled the controls for the potentially explosive gas mixtures contained in the system will not be required. There will no longer be any source of explosive gas generated from reactor operation.

Tanks for storing radioactive liquid waste will be maintained after shutdown. Therefore, the description is revised to reflect a Storage Tank Radioactivity Monitoring Program. The formatting for this specification is proposed to be modified to reflect the elimination of the discussion pertaining to the concentration of hydrogen gas in the Offgas System.

Page 74 of 85

Attachment 1 Evaluation of Proposed Changes TS Section 5.5.9 - Diesel The Diesel Fuel Oil Testing Program was established to implement the Fuel Oil Testing Program required testing of both new and stored fuel oil for the EDGs. This program is proposed for elimination from the PDTS since the EDGs will not perform any safety function in the permanently shutdown and defueled facility. The TS associated with the EDGs and the diesel fuel oil subsystem (TS 3.8.1, 3.8.2, and 3.8.3) are proposed for removal from the PDTS as described above.

TS 5.5.10 - Technical The Technical Specifications Bases Control Program is being modified Specifications (TS) Bases to reflect that once the facility is permanently defueled the title of the Control Program UFSAR will be revised to DSAR.

TS Section 5.5.11 - Safety This program was established to ensure loss of safety function is Function Determination detected and appropriate actions taken. The SFDP is proposed for Program (SFDP) 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 Section 5.5.12 - Primary This program was established to implement the leakage rate testing of Containment Leakage Rate the primary containment as required by 10 CFR 50.54(o) and Testing Program 10 CFR 50 Appendix J, Option B, as modified by exemptions. This program will not be retained in the PDTS, because the Primary Containment Leakage Rate Testing Program pertains only to reactor support systems that are not needed in a permanently defueled condition. The requirements in TS 3.6.1.1, 3.6.1.2, and 3.6.1.3, for primary containment systems are being deleted as described above.

Therefore, this specification does not apply in the permanently shutdown and defueled condition.

TS Section 5.5.13 - Battery This program was established to provide for Station battery restoration Monitoring and Maintenance and maintenance. As discussed above, TS Section 3.8, including all Program requirements for DC sources and battery parameters are proposed for deletion in their entirety. Therefore, the requirements for the maintenance, testing, and replacement of station batteries described in this specification are similarly unnecessary and are proposed for deletion following the establishment of a permanently shutdown and defueled condition for Dresden.

TS Section 5.5.14 - Control This program was established and implemented to ensure that the Room Envelope Habitability Control Room Envelope (CRE) habitability was maintained such that, Program with an operable CREV 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.(22) and 2.EE for Dresden, Units 2 and 3, 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 Dresden are located outside the control room. As such, the need for the operator to occupy the CRE is diminished Page 75 of 85

Attachment 1 Evaluation of Proposed Changes following the establishment of the permanent shutdown and defueled condition. Additionally, as previously discussed, TS 3.3.7.1 and 3.7.4 are not proposed to be included in the PDTS; thus, Technical Specification 5.5.13 is proposed for deletion.

TS Section 5.5.15 - This program provides controls for Surveillance Frequencies. The Surveillance Frequency program shall ensure that SRs specified in the TS are performed at Control Program intervals sufficient to assure the associated LCOs are met.

The requirements regarding the Surveillance Frequency Control Program (SFCP) are proposed for deletion. The Technical Specifications proposed for retention in the PDTS contain only one SR.

Therefore, there is no need to maintain this program to control the Frequency of this one SR and it can be eliminated. Thus, Technical Specification 5.5.15 is proposed for deletion.

TS Section 5.6.2 - Annual This reporting requirement is being retained in the PDTS with minor Radiological Environmental editorial changes. The NOTE regarding the ability to make a single Operating Report submittal for a multiple unit station is proposed for deletion. Once Dresden has ceased operations, Units 1, 2, and 3, will be referred to as the facility. Additionally, the term "unit" is replaced with the term "facility," because the term "unit" generally refers to an operating reactor, and since the reactors can no longer be operated following the submittal of the required certifications in accordance with 10 CFR 50.82(a)(2), whereas the term "facility" more appropriately refers to the overall site.

TS Section 5.6.3 - This reporting requirement is being retained in the PDTS with minor Radioactive Effluent Release editorial changes. The NOTE regarding the ability to make a single Report submittal for a multiple unit station is proposed for deletion. Once Dresden has ceased operations, Units 1, 2, and 3, will be referred to as the facility. Additionally, the term "unit" is replaced with the term "facility," because the term "unit" generally refers to an operating reactor, and since the reactors can no longer be operated following the submittal of the required certifications in accordance with 10 CFR 50.82(a)(2), whereas the term "facility" more appropriately refers to the overall site.

TS Section 5.6.5 - Core According to TS Section 5.6.5, the core operating limits shall be Operating Limits Report established prior to each reload cycle, or prior to any remaining portion (COLR) of a reload cycle and 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 PDTS, because the Dresden, 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 Section 5.6.6 - Post This report is required by Condition B or F of LCO 3.3.3.1. The report Accident Monitoring (PAM) outlines the preplanned alternate method of monitoring, the cause of Instrumentation Report inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to an OPERABLE status. This reporting requirement will not be retained in the PDTS because the Dresden, 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 Page 76 of 85

Attachment 1 Evaluation of Proposed Changes the NRC. As discussed above, the requirements in TS 3.3.3.1 regarding PAM instrumentation are proposed to be deleted: therefore, the PAM instrument report will not be required in the permanently shutdown and defueled condition and is proposed for deletion. provides the existing TS pages for Dresden, 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 Dresden on or before November 30, 2021.

Once the certifications for permanent cessation of operations and removal of fuel from the reactor vessels are submitted for Dresden, Units 2 and 3, to the NRC pursuant to 10 CFR 50.82(a)(1)(i) and (ii), NRC regulations stipulated in 10 CFR 50.82(a)(2) will no longer authorize operation of the Units 2 and 3, reactors or emplacement or retention of fuel in the reactor vessels under the 10 CFR 50 licenses. As a result, Dresden 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 Page 77 of 85

Attachment 1 Evaluation of Proposed Changes 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 Dresden 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 DBA 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 Dresden 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 DBA 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 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 Dresden, 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 Page 78 of 85

Attachment 1 Evaluation of Proposed Changes will be applicable to Dresden 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 Dresden 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 organization 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 Dresden 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 Dresden 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 Reactors 10 CFR 50.46(a)(1)(i) states 'This section does not apply to a nuclear power reactor facility for which the certifications required under §50.82(a)(1) have been submitted."

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; Page 79 of 85

Attachment 1 Evaluation of Proposed Changes (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 10 CFR 50.62(a) 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."

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 Page 80 of 85

Attachment 1 Evaluation of Proposed Changes 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 NRC 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."

Design Basis Accidents (DBAs)

Chapter 15 of the Dresden UFSAR describes the DBA 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 NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii) for Dresden Units 2 and 3, the 10 CFR 50 licenses will no longer permit operation of the reactors or placement 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 UFSAR Chapter 15 will no longer be applicable after Dresden is in the permanently defueled condition. The events that remain applicable to Dresden, Units 2 and 3, in the permanently shutdown and defueled condition are the Fuel Handling Accident (FHA) within the SFPs and Postulated Radioactive Release Due to Liquid Tank Failure.

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.

DPR-19 and DPR-25 for Dresden Nuclear Power Station, Units 2 and 3 (Dresden). 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 Dresden Permanently Defueled Technical Specifications (PDTS). 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)(i) and (ii), NRC regulations stipulated in Page 81 of 85

Attachment 1 Evaluation of Proposed Changes 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 Dresden has permanently ceased operation and have entered a permanently defueled condition. Because the 10 CFR 50 licenses for Dresden 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 is no longer credible.

Dresden's accident analyses are described in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). The remaining postulated design basis accident events that could potentially occur at Dresden in a permanently defueled condition would be a postulated liquid release due to liquid tank failure, and a fuel handling accident (FHA) in the SFP. The FHA analyses for Dresden shows that after 48 days of decay time after the reactors have shutdown and provided the SFP water level requirement of TS LCO 3.7.8 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 certifications of permanent removal of fuel from the reactor vessels until 48 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.

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 are 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 Dresden facility has no impact on the remaining applicable DBAs.

Page 82 of 85

Attachment 1 Evaluation of Proposed Changes The removal of LCOs or 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 DBAs previously evaluated since these DBAs will no longer be applicable in the permanently defueled condition.

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

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

Response: No 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 only related to the operation of the nuclear reactors, or 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 Dresden, Units 2 and 3, RFOLs, additional conditions, and TS do not affect systems credited in the accident analysis for the remaining DBAs. The proposed RFOLs and PDTS will continue to require proper control and monitoring of safety significant parameter and activities. The TS regarding SFP water level is 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.

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 involve deleting or modifying the RFOLs, additional conditions, and TS once the Dresden facility has been permanently shutdown and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 licenses for Dresden will no longer authorize operation of the reactors or emplacement or retention of fuel into the reactor vessels Page 83 of 85

Attachment 1 Evaluation of Proposed Changes following submittal of the certifications required by 10 CFR 50.82(a)(1). Because the 10 CFR 50 licenses for Dresden 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 DBAs that could potentially occur at Dresden in a permanently defueled condition are the FHA and the Postulated Radioactive Release Due to Liquid Radwaste Tank Failures. The proposed changes do not adversely affect the inputs or assumptions of the 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 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 the Dresden 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 certifications of permanent removal of fuel from the reactor vessels until 48 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 amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

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

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 amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, Page 84 of 85

Attachment 1 Evaluation of Proposed Changes the proposed amendments meet the eligibility criterion for categorical exclusion set for 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) to NRC, "Certification of Permanent Cessation of Power Operations for Dresden Nuclear Power Station, Units 2 and 3," dated September 2, 2020 (ADAMS Accession No. ML20246G627)

2. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Request for Approval of Certified Fuel Handler Training Program,"

for Byron Station, Units 1 and 2, and Dresden Nuclear Power Station, Units 1, 2 and 3, dated September 24, 2020 (ADAMS Accession No. ML20269A233)

3. Letter from Patrick R. Simpson (Exelon to NRC, " License Amendment Request - Proposed Changes to Unit 1 Technical Specifications Sections 6.1, "Responsibility," and Units 2 and 3 Technical Specifications 1.1, "Definitions," and 5.0, "Administrative Controls," for Permanently Defueled Condition," dated September 24, 2020 (ADAMS Accession No. ML20269A404)

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 Alternate Source Terms," March 7, 2006

6. NUREG-1796, "Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2," dated October 2004 (ADAMS Accession No. ML043060582)

Page 85 of 85

Attachment 2 Markup of Renewed Facility Operating License Pages Dresden Nuclear Power Station Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 MARKED UP UNIT 2 FACILITY OPERATING LICENSE PAGES 1-11 Appendix B MARKED UP UNIT 3 FACILITY OPERATING LICENSE PAGES 1-11 Appendix B, Pages 1 and 2

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-19

1. The U.S. Nuclear Regulatory Commission (Commission) having previously made the findings set forth in license No. DPR-19 issued on February 20, 1991, has now found that: --

A. The application to renew license No. DPR-19 filed by the Exelon Generation Company, LLC* (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I, and all required notifications Deleted; to other agencies or bodies have been duly made; B. Construction of the Dresden Nuclear Power Station, Unit 2 (the facility) has been completed in conformity with Construction Permit No. CPPR-18 and the application, as amended, the provisions of the Act, and the regulations of the Commission, and has been operating under a provisional license since December 22, 1969; C. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54,3, for Dresden Nuclear Power Station, Unit 2 (facility or plant), and that any changes made to the plant's current licensing basis in order to comply with 10 CFR 54.29(a) are in accord with the Act and the Commission's regulations; be maintained D. The facility will operate 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);

• 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. DPR-19 Amendment No.

E. There is reasonable assurance: (i) that the activities authorized by this renewed operating 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 2.0 below);

F. Exelon Generation Company, LLC, is technically qualified to engage in the activities authorized by this renewed operating license in accordance with the rules and regulations of the Commission; G. Exelon Generation Company has satisfied the applicable provisions of 10 CFR Part 140, uFinancial Protection Requirements and Indemnity Agreements," of the Commission's regulations; H. The issuance of this renewed operating license will not be inimical to the common defense* and security or to the health and safety of the public; I. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this Renewed Facility Operating License No. DPR-19 is in accordance with 10 CFR Part 51 of the Commission's Deleted. regulations and all applicable requirements have been satisfied; and J. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed operating license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.

2. On the basis of the foregoing findings regarding this facility, Facility Operating License No. DPR-19, issued February 20, 1991, is superseded by Renewed Facility Operating License No. DPR-19, which is hereby issued to Exelon Generation Company, LLC, to read as follows:

permanently A. This renewed operating license applies to the Dresden Nuclear Power Station, defueled Unit 2, a boiling water reactor and associated equipment (the facility). The facility is located in Grundy 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: as required for irradiated fuel storage and use (1) Exelon Generation Company, LLC, pursuant to Section 104b of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated location in Grundy County, Illinois, in accordance with the procedures and limitations set forth in this renewed operating license; Renewed License No. DPR-19 Amendment No.

that were used (2) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment or calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.

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

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core Deleted. power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications Permanently The Technical Specifications contained in Appendix A, as revised through Defueled Amendment No. 270, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

maintain Permanently Defueled (3) Operation in the coastdown mode is permitted to 40% power.

Deleted.

to possess any byproduct, source and special nuclear material such as sealed neutron sources previously used for reactor startup or reactor instrumentation, and fission detectors; Renewed License No. DPR-19 Amendment No. 270

(4) The valves in the equalizer piping between the recirculation loop shall be Deleted.

closed at all times during reactor operation.

(5) The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

Deleted.

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

Deleted.

(6) Surveillance Requirements The Surveillance Requirements contained in Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 150:

a. Surveillance Requirement 4.1.A.2 - RPS Logic System Functional Test

b. Surveillance Requirement 4.2.A.2 - Primary & Secondary Containment Logic System Functional Test

c. Surveillance Requirement 4.2.J.2 - Feedwater Pump Trip Logic System Functional Test

d. Surveillance Requirement 4.6.F.1.b - Relief Valve Logic System Functional Test

e. Surveillance Requirement 4.9.A.9 - Simultaneous Diesel Generator Start

f. Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fifteenth refueling outage (D2R15).

Renewed License No. DPR-19 Amendment No.

(7) Additional Conditions Deleted. The Additional Conditions contained in Appendix B, as revised through Amendment No. 191, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Additional Conditions.

(8) Deleted (9) Deleted (10) 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, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company, LLCs consolidated net utility plant, as recorded on Exelon Generation Company, LLCs books of account.

(11) Deleted.

(12) Deleted.

(13) Deleted.

(14) Exelon Generation Company, LLC shall relocate certain Technical Specification requirements to EGG-controlled documents upon Deleted. implementation of the Amendment No. 185. The items and appropriate documents are as described in Table LA, Removal of Details Matrix, and Table R, Relocated Specifications, that are attached to the NRCs Safety Evaluation enclosed with Amendment No. 185.

Renewed License No. DPR-19 Amendment No. 267

(15) The schedule for performing Surveillance Requirements (SRs) that are Deleted. new or revised in Amendment No. 185 shall be as follows:

For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of Amendment No. 185.

For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of Amendment No. 185.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of Amendment No. 185.

For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to implementation of Amendment No. 185.

(16) Following implementation of Amendment No. 185, the reactor protection Deleted. system trip setpoint for main steam isolation valve closure shall be maintained at the previous setpoint (less than or equal to 10% closed) until startup after the first outage of sufficient duration to change the setpoint.

(17) The license is amended to authorize changing the UFSAR to allow credit for containment overpressure as detailed below, to assure adequate Net Deleted. Positive Suction Head is available for low pressure Emergency Core Cooling System pumps following a design-basis accident.

From (sec) To (sec) Credit (psig)

Accident start 290 9.5 290 5,000 4.8 5,000 30,000 6.6 30,000 40,000 6.0 40,000 45,500 5.4 45,500 52,500 4.9 52,500 60,500 4.4 60,500 70,000 3.8 70,000 84,000 3.2 84,000 104,000 2.5 104,000 136,000 1.8 136,000 Accident end 1.1 Renewed License No. DPR-19 Amendment No. 267

(18) 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 (19) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

Renewed License No. DPR-19 Amendment No. 267

Deleted.

(20) Upon implementation of Amendment No. 226 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 6 months if not performed previously.

(21) Upon implementation of Amendment No. 249 the licensee shall adhere to the following requirements as part of the DNPS unit 2 spent fuel pool coupon surveillance program to ensure that the B-10 areal density of the BORAL remains at or above its minimum credited value and that the regulatory requirement to maintain the Technical Specification value of keff 0.95 continues to be met:

1. Ensure that coupon measurements of B-10 areal density are performed by a qualified laboratory;

2. Ensure that the coupons are removed for evaluation every 10 years;

3. Ensure that should any coupon be identified as failing the minimum certified B-10 areal density criterion based on coupon test results, EGC will perform in-situ testing to confirm that the minimum B-10 areal density (0.02 g/cm2) is met for the BORAL panels installed in the DNPS spent fuel pools; and, Renewed License No. DPR-19 Amendment No. 267

4. Submit a report to the NRC within 90 days following the completion of evaluations associated with Item 3 above. The report shall include; a description of the testing results, the assessments performed, and the interim and long-term corrective actions for abnormal indications.

Deleted.

D. The facility has been granted certain exemptions from the requirements of Section III.G of Appendix R to 10 CFR Part 50, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. These exemptions were granted and sent to the licensee in letters dated February 2, 1983, September 28, 1987, July 6, 1989, and August 15, 1989.

In addition, the facility has been granted certain exemptions from Sections II and III of Appendix J to 10 CFR Part 50, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. This section contains leakage test requirements, schedules and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted and sent to the licensee in a letter dated June 25, 1982.

These exemptions granted pursuant to 10 CFR 50.12 are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

Deleted.

E. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated March 22, 1978 with supplements dated December 2, 1980, and February 12, 1981; January 19, 1983; July 17, 1987; September 28, 1987; and January 5, 1989, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

F. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements (22) Handling of irradiated fuel in the spent fuel pool will not be permitted following implementation of the PDTS until a minimum of 48 days following permanent shutdown.

Renewed License No. DPR-19 Amendment No. 267

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: "Dresden Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," 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. 238 as modified by License Amendment No. 246.

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.

Deleted I. Updated Final Safety Analysis Report The Exelon Generation Company, LLC Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. The Exelon Generation Company, LLC shall complete these activities no later than December 22, 2009, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e}(4} following issuance of this renewed license. Until that update is complete, Exelon Generation Company, LLC may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Exelon Generation Company, LLC evaluates such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Deleted J. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

1The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-19 Amendment No. 246

3. This renewed operating license is effective as of the date of issuance and shall expire at midnight on December 22, 2029.

FOR THE NUCLEAR REGULATORY COMMISSION is effective until the Commission

/RA/

notifies the licensee in writing J. E. Dyer, Director that the license is Office of Nuclear Reactor Regulation terminated.

Attachments:

1. Appendix A - Technical Specifications

2. Appendix B - Additional Conditions Deleted Date of Issuance: October 28, 2004 Renewed License No. DPR-19 Amendment No. 226

APPENDIX B ADDITIONAL CONDITIONS Deleted FACILITY OPERATING LICENSE NO. DPR-19 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date 157 The EOPs shall be changed to alert Shall be operator to NPSH concerns and to make implemented within containment spray operation consistent 30 days after with the overpressure requirements for issuance of NPSH. Amendment No. 157.

160 This amendment authorizes the licensee to 30 days from the incorporate in the Updated Final Safety date of issuance Analysis Report (UFSAR}, the description of Amendment of the Reactor Coolant System design No. 160.

pressure, temperature and volume that was removed from Technical Specification Section 5.4, and evaluated in a safety evaluatjon dated June 12, 1997.

163 The licensee shall review the Dresden 60 days from the Operation Annunciator and General Abnormal date of issuance Conditions Procedures and revise them as of Amendment required to ensure operator action is taken No. 163 in a timely manner to limit occupational doses and environmental releases.

Amendment No. 191

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-25 The U.S. Nuclear Regulatory Commission (Commission) having previously made the findings set forth in License No. DPR-25 issued on January 12, 1971, has now found that:

a. The application to renew License No. DPR-25 filed by the Exelon Generation Company, LLC (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made;

b. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for Dresden Nuclear Power Station, Unit 3

. (facility or plant), and that any changes made to the plant's current licensing basis in order to comply with 10 CFR 54.29(a) are in accord with the Act and the Commission's regulations;

c. The applicant* has submitted to the Commission all technical information required by Provisional Construction Permit No. CPPR-22, the Atomic Energy Act of 1954, as amended (the Act), and the rules and regulations of the Commission to complete the application for a construction permit and facility license dated February 10, 1966, as supplemented by application for a facility license dated November 17, 1967 and amended by Amendment Nos. 8 through 24, dated August 30, 1968, November 21, 1968, February 28, 1969, March 18, 1969, April 16, 1969, May 20, 1969, July 2, 1969, July 22, 1969, Augusts, 1969,August8, 1969,August10, 1969,August18, 1969, September 2, 1969, October 16, 1969, May 7, 1970, August 11, 1970 and September 4, 1970, respectively, (the application); and supplemented by the applicant's letter dated December 17, 1970, and telegram dated December 18, 1970;

• 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. DPR-25 Amendment No.

Deleted; d. The Dresden Nuclear Power Station Unit 3 (the facility) has been substantially completed in conformity with Provisional Construction Permit No. CPPR-22, the application, the provisions of the Act and the rules and regulations of the Commission; be maintained

e. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; maintained

f. There is reasonable assurance: (i) that the facility can be operated at power levels not in excess of 2957 megawatts (thermal) in accordance with this license without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission;

g. Exelon Generation Company, LLC is technically and financially qualified to engage in the activities authorized by this renewed operating license in accordance with the rules and regulations of the Commission;

h. Exelon Generation Company, LLC has furnished proof of financial protection to satisfy the requirements of 10 CFR Part 140;

i. The issuance of this renewed operating license will not be inimical to the common defense and security or to the health and safety of the public; and

j. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this Renewed Facility Operating License No.

DPR-25 is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

On the basis of the foregoing findings regarding this facility, Facility Operating License No. DPR-25, issued January 12, 1971, is superseded by Renewed Facility Operating License No. DPR-25, which is hereby issued to Exelon Generation Company, LLC (EGC or the licensee), to read as follows:

1. This renewed operating license applies to the Dresden Nuclear Power Station, Unit 3, a single cycle, boiling, light water reactor and electric generating equipment (the facility).

The facility is located at the Dresden Nuclear Power Station in Grundy County, Illinois, and is described in the "Safety Analysis Report," as supplemented and amended permanently (Amendment Nos. 8 through 24).

defueled,

2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Exelon Generation Company, LLC: and use A. Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use, and operate the facility as Renewed License No. DPR-25 Amendment No.

required for irradiated fuel storage a utilization facility at the designated location at the Dresden Nuclear Power Station, in accordance with the procedures and limitations set forth in this renewed operating license; that was used to possess any B. Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time byproduct, special nuclear material, not including plutonium, as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as source and described inthe Final Safety Analysis Report, as supplemented and amended as special nuclear of September 3, 1976; material such as or sealed neutron C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear materials as sealed sources neutron sources for reactor startup, sealed sources for reactor instrumentation previously used and radiation monitoring equipment calibration, and as fission detectors in for reactor amounts required; startup or reactor D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and instrumentation, use in amounts as required any byproduct, source or special nuclear material and fission without restriction to chemical or physical form, for sample analysis or instrument detectors; and equipment calibration or associated with radioactive apparatus or components; and that were E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Dresden Nuclear Power Station, Unit Nos. 1, 2, and 3.

Deleted.

F. Surveillance Requirements The Surveillance Requirements contained in Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 145:

a. Surveillance Requirement 4.1.A.2 - RPS Logic System Functional Test

b. Surveillance Requirement 4.2.A.2 - Primary & Secondary Containment Logic System Functional Test

c. Surveillance Requirement 4.2.J.2 - Feedwater Pump Trip Logic System Functional Test

d. Surveillance Requirement 4.6.F.1.b - Relief Valve Logic System Functional Test

e. Surveillance Requirement 4.9.A.9 - Simultaneous Diesel Generator Start Renewed License No. DPR-25 Amendment No.

f. Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14).

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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:

Deleted.

A. Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 2957 megawatts (thermal), except that the licensee shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in the licensees telegrams; dated February 26, 1971, have been verified in writing by the Commission.

B. Technical Specifications Permanently The Technical Specifications contained in Appendix A, as revised through Defueled Amendment No. 263, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

maintain C. Reports Permanently Defueled The licensee shall make certain reports in accordance with the requirements of Permanently the Technical Specifications.

Defueled D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

Deleted. Permanently Defueled E. Restrictions Operation in the coastdown mode is permitted to 40% power.

Renewed License No. DPR-25 Amendment No. 263

Deleted.

F. The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRG approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRG staff for review and approval prior to implementation of the change.

Deleted.

G. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated March 22, 1978 with supplements dated December 2, 1980, and February 12, 1981; January 19, 1983; July 17, 1987; September 28, 1987; and January 5, 1989, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

H. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, 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 plans\ which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Dresden Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," 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. 231 and modified by License Amendment No. 239.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-25 Amendment No. 239

- Sa -

I. Deleted J. Deleted K. Deleted L. Deleted [Arndt. 87, 7-24-86)

Renewed License No. DPR-25 Amendment No. 231

M. Deleted [Arndt. 85, 12-12-85]

Deleted.

N. By Amendment No. 144, the license is amended to allow, on a one time temporary basis, operation of Dresden, Unit 3, with the corner room structural steel members in the Low Pressure Coolant Injection Corner Rooms outside the Updated Final Safety Analysis Report (UFSAR) design parameters. Operation under these conditions is allowed up to and including the next scheduled refueling outage (D3R14).

The repairs to Dresden, Unit 3, corner room structural steel shall restore the steel design margins to the current UFSAR (updated through Revision 1A) design criteria. The design of the modifications to the Dresden, Unit 3, corner room structural steel members will be based on use of elastic section modulus and the structural steel stresses will be limited to 1.6 of the American Institute of Steel Construction (AISC allowables). The modifications to Dresden, Unit 3, corner room structural steel will be implemented during the upcoming D3R14 refueling outage.

During this interim period of operation, should vibratory ground motion exceeding the UFSAR Operating Basis Earthquake (OBE) design parameters, Dresden, Unit 3, will be shut down for inspection and will not start up without prior NRC approval.

0. Additional Conditions Deleted. The Additional Conditions contained in Appendix B, as revised through Amendment No. 185, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Additional Conditions.

P. Deleted Q. Deleted R. 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, facilities 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.

Renewed License No. DPR-25 Amendment No. 227

S. Deleted.

T. Deleted.

U. Deleted.

Deleted.

V. Exelon Generation Company, LLC shall relocate certain Technical Specification requirements to EGC-controlled documents upon implementation of the Amendment No. 180. The items and appropriate documents are as described in Table LA, Removal of Details Matrix, and Table R, Relocated Specifications, that are attached to the NRCs Safety Evaluation enclosed with Amendment No. 180.

Deleted.

W. The schedule for performing Surveillance Requirements (SRs) that are new or revised in Amendment No. 180 shall be as follows:

For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of Amendment No. 180.

For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of Amendment No. 180.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of Amendment No. 180.

For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to implementation of Amendment No. 180.

Renewed License No. DPR-25 Amendment No. 260

Deleted.

X. The license is amended to authorize changing the UFSAR to allow credit for containment over pressure as detailed below, to assure adequate Net Positive Suction Head is available for low pressure Emergency Core Cooling System pumps following a design-basis accident.

From (sec) To (sec) Credit (psig)

Accident start 290 9.5 290 5,000 4.8 5,000 30,000 6.6 30,000 40,000 6.0 40,000 45,500 5.4 45,500 52,500 4.9 52,500 60,500 4.4 60,500 70,000 3.8 70,000 84,000 3.2 84,000 104,000 2.5 104,000 136,000 1.8 136,000 Accident end 1.1 Renewed License No. DPR-25 Amendment No. 260

Deleted.

Y. Updated Final Safety Analysis Report The Exelon Generation Company, LLC Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. The Exelon Generation Company, LLC shall complete these activities no later than January 12, 2011, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, Exelon Generation Company, LLC may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Exelon Generation Company, LLC evaluates such change pursuant to the criteria set forth in Deleted. 10 CFR 50.59 and otherwise complies with the requirements in that section.

Z. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

AA. 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. DPR-25 Revised by letter dated August 23, 2007 Amendment No.

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

1. Water spray scrubbing

2. Dose to onsite responders BB. The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

Deleted.

CC. Upon implementation of Amendment No. 218 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered met.

Following implementation:

(1) The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(2) The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from January 1997, the date of the most recent successful tracer gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(3) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 6 months if not performed previously.

Renewed License No. DPR-25 Amendment No. 218

-10A-DD. Upon implementation of Amendment No. 242 the licensee shall adhere to the following requirements as part of the DNPS unit 3 spent fuel pool coupon surveillance program to ensure that the B-1 O areal density of the BORAL remains at or above its minimum credited value and that the regulatory requirement to maintain the Technical Specification value of kett::;, 0.95 continues to be met:

(1) Ensure that coupon measurements of B-1 O areal density are performed by a qualified laboratory; (2) Ensure that the coupons are removed for evaluation every 1O years; (3) Ensure that should any coupon be identified as failing the minimum certified B-10 areal density criterion based on coupon test results, EGC will perform in-situ testing to confirm that the minimum B-1 O areal density (0.02 g/cm 2) is met for the BORAL panels installed in the DNPS spent fuel pools; and, (4) Submit a report to the NRC within 90 days following the completion of evaluations associated with Item 3 above. The report shall include; a description of the testing results, the assessments performed, and the interim and long-term corrective actions for abnormal indications.

EE. Handling of irradiated fuel in the spent fuel pool will not be permitted following implementation of the PDTS until a minimum of 48 days following permanent shutdown.

Renewed License No. DPR-25 Amendment No. 242

4. This renewed operating license is effective as of the date of issuance and shall expire at midnight on January 12, 2031.

FOR THE NUCLEAR REGULATORY COMMISSION is effective until the Commission /RA/

notifies the licensee in writing J. E. Dyer, Director that the license is Office of Nuclear Reactor Regulation terminated.

Attachments:

1. Appendix A - Technical Specifications

2. Appendix B - Additional Conditions Deleted Date of Issuance: October 28, 2004 Renewed License No. DPR-25 Amendment No. 218

APPENDIX B ADDITIONAL CONDITIONS Deleted FACILITY OPERATING LICENSE NO. DPR-25 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date 152 The licensee shall complete the evaluation Prior to Unit 3 of the torus attached piping. returning to Mode 3 from refueling outage D3R14.

152 The EOPs shall be changed to alert Shall be operator to NPSH concerns and to make implemented within containment spray operation consistent 30 days after with the overpressure requirements for issuance of NPSH. Amendment No. 152.

155 This amendment authorizes the licensee to 30 days from the incorporate in the Updated Final Safety date of issuance Analysis Report (UFSAR), the description of Amendment of the Reactor Coolant System design No. 155.

pressure, temperature and volume that was removed from Technical Specification Section 5.4, and evaluated in a safety evaluation dated June 12, 1997.

v- Amendment No. 185

Amendment Implementation Number Additional Condition Date 158 The licensee shall review the Dresden 60 days from the Operation Annunciator and General Abnormal date of issuance Conditions Procedures and revise them as of Amendment No. 158 required to ensure operator action is taken in a timely manner to limit occupational doses and environmental releases.

158 The licensee shall change the set points for 60 days from the the Main Steam Line Radiation Monitor and date of issuance Offgas System Radiation Monitor alarms to of Amendment No. 158 1.5 times the normal full power N-16 background {with hydrogen addition) dose rates.

Amendment No. 158

Attachment 3 Markup of Technical Specifications Pages Dresden Nuclear Power Station Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 MARKED UP UNITS 2 AND 3 TECHNICAL SPECIFICATIONS PAGES 1.1-1 through 1.1-8 1.3-1 through 1.3-13 1.4-1 through 1.4-5 3.0-1 through 3.0-5 3.7.8-1 4.0-1 5.1-1 5.2-1 5.2-2 5.3-1 5.4-1 5.5-2 through 5.5-14 5.6-1 through 5.6-6

Definitions 1.1 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.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR) LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass CERTIFIED FUEL all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.

HANDLER* Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

(continued)

A CERTIFIED FUEL HANDLER is an individual who complies with the provisions of the CERTIFIED FUEL HANDLER Training Program required by Specification 5.3.2.*

• TS Markups incorporate changes from Attachment 2 of Dresden Administrative Controls LAR submitted on September 24, 2020 (ML20269A404)

Dresden 2 and 3 1.1-1 Amendment No. 266/259

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

(continued)

Dresden 2 and 3 1.1-2 Amendment No. 266/259

Definitions 1.1 1.1 Definitions (continued)

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

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

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be the inhalation committed dose conversion factors in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

(continued)

Dresden 2 and 3 1.1-3 Amendment No. 266/259

Definitions 1.1 1.1 Definitions (continued)

DRAIN TIME The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:

a. The water inventory above the TAF is divided by the limiting drain rate;

b. The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error),

for all penetration flow paths below the TAF except:

1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

(continued)

Dresden 2 and 3 1.1-4 Amendment No. 266/259

Definitions 1.1 1.1 Definitions DRAIN TIME c. The penetration flow paths required to be (continued) evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory;

d. No additional draining events occur; and

e. Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

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

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; (continued)

Dresden 2 and 3 1.1-5 Amendment No. 266/259

Definitions

• TS Markups incorporate changes from Attachment 2 1.1 of Dresden Administrative Controls LAR submitted on 1.1 Definitions September 24, 2020 (ML20269A404)

LEAKAGE c. Total LEAKAGE (continued)

Sum of the identified and unidentified LEAKAGE; and

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

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

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the NON-CERTIFIED appropriate correlation(s) to cause some point in OPERATOR* the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLEOPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, A NON-CERTIFIED OPERATOR is a non-licensed operator (continued) who complies with the qualification requirements of Specification 5.3.1, but is not a CERTIFIED FUEL HANDLER.*

Dresden 2 and 3 1.1-6 Amendment No. 266/259

Definitions 1.1 1.1 Definitions OPERABLEOPERABILITY lubrication, and other auxiliary equipment that (continued) are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2957 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from the opening of the sensor contact until the TIME opening of the trip actuator. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

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

a. The reactor is xenon free;

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

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

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

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME shall be RESPONSE TIME that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Dresden 2 and 3 1.1-7 Amendment No. 266/259

Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE

(°F) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot NA Standby 3 Hot Shutdown(a) Shutdown 212 4 Cold Shutdown(a) Shutdown 212 5 Refueling(b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

Dresden 2 and 3 1.1-8 Amendment No. 266/259

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.

BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that storage and typically describe the ways in which the requirements of the handling of LCO can fail to be met. Specified with each stated irradiated fuel 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 facility unit is in a MODE or specified condition stated in the Applicability of the LCO. Unless otherwise specified, the Completion Time begins when a senior licensed operator on the operating shift crew with responsibility for plant operations makes the determination that an LCO is not met and an ACTIONS Condition is entered. The "otherwise specified" exceptions are varied, such as a Required Action Note or Surveillance Requirement Note that provides an alternative time to perform specific tasks, such as testing, without starting the Completion Time. While utilizing the Note, should a Condition be applicable for any reason not addressed by the Note, the Completion Time begins. Should the time allowance 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 3.8.1, "AC Sources - Operating," Required Action B.2, requires declaring required feature(s) supported by an inoperable diesel generator, inoperable when the redundant required feature(s) are inoperable. The Completion Time states, "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery of Condition B concurrent with inoperability of redundant required feature(s)." In this case the Completion Time does not begin until 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 (continued)

Dresden 2 and 3 1.3-1 Amendment No. 255/248

Completion Times 1.3 1.3 Completion Times DESCRIPTION longer exists or the unit is not within the LCO (continued) Applicability.

facility If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the discovery of the situation that required entry into the Condition, unless otherwise specified.

Once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition, unless otherwise specified.

However, when a subsequent division, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concurrent with the first inoperability; and

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extension does not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each division, (continued)

Dresden 2 and 3 1.3-2 Amendment No. 255/248

Completion Times 1.3 1.3 Completion Times DESCRIPTION subsystem, component or variable expressed in the Condition)

(continued) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e.,

"once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ."

EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.

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 B are to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for reaching MODE 3 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) is allowed for reaching MODE 4 from (continued)

Dresden 2 and 3 1.3-3 Amendment No. 262/255

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-1 (continued) the time that Condition B was entered. If MODE 3 is reached within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the time allowed for reaching MODE 4 is the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> because the total time allowed for reaching MODE 4 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

If Condition B is entered while in MODE 3, the time allowed for reaching MODE 4 is the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE status.

B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.

When a pump is declared inoperable, Condition A is entered.

If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump.

The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.

(continued)

Dresden 2 and 3 1.3-4 Amendment No. 255/248

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.

On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for 7 days.

(continued)

Dresden 2 and 3 1.3-5 Amendment No. 255/248

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7 days Function X Function X subsystem subsystem to inoperable. OPERABLE status.

B. One B.1 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Function Y Function Y subsystem subsystem to inoperable. OPERABLE status.

C. One C.1 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Function X Function X subsystem subsystem to inoperable. OPERABLE status.

AND OR One C.2 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Function Y Function Y subsystem subsystem to inoperable. OPERABLE status.

(continued)

Dresden 2 and 3 1.3-6 Amendment No. 262/255

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

When one Function X subsystem and one Function Y subsystem are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each subsystem, starting from the time each subsystem was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second subsystem was declared inoperable (i.e., the time the situation described in Condition C was discovered).

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected subsystem was declared inoperable (i.e., initial entry into Condition A).

It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended.

(continued)

Dresden 2 and 3 1.3-7 Amendment No. 262/255

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-4 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves to OPERABLE inoperable. status.

B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.

A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis.

Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.

Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided this does not result in any subsequent valve being inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If the Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (plus the extension) expires while one or more valves are still inoperable, Condition B is entered.

(continued)

Dresden 2 and 3 1.3-8 Amendment No. 185/180

Completion Times 1.3 1.3 Completion Times EXAMPLE EXAMPLE 1.3-5 (continued)

ACTIONS

---

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

Separate Condition entry is allowed for each inoperable valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves OPERABLE status.

inoperable.

B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.

The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.

The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve.

(continued)

Dresden 2 and 3 1.3-9 Amendment No. 185/180

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued)

If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve.

If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve.

Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply.

EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform Once per inoperable. SR 3.x.x.x. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> OR A.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated Completion Time not met.

(continued)

Dresden 2 and 3 1.3-10 Amendment No. 185/180

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-6 (continued)

Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance.

The initial 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be completed within the first 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval. If Required Action A.1 is followed and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered.

If Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is not met, Condition B is entered.

If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.

(continued)

Dresden 2 and 3 1.3-11 Amendment No. 185/180

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> subsystem subsystem inoperable. isolated. AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND A.2 Restore subsystem 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to OPERABLE status.

B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.

Required Action A.1 has two Completion Times. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time begins at the time the Condition is entered and each "Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter" interval begins upon performance of Required Action A.1.

If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or any subsequent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 (continued)

Dresden 2 and 3 1.3-12 Amendment No. 185/180

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.

IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.

Dresden 2 and 3 1.3-13 Amendment No. 185/180

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated Limiting Condition for Operation (LCO). An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.

Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both. Example 1.4-4 discusses these special situations.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria. SR 3.0.4 restrictions would not apply if both the following conditions are satisfied:

(continued)

Dresden 2 and 3 1.4-1 Amendment No. 185/180

Frequency 1.4 1.4 Frequency DESCRIPTION a. The Surveillance is not required to be performed; and (continued)

b. The Surveillance is not required to be met or, even if required to be met, is not known to be failed.

illustrates the manner in which EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3. The examples do not reflect the potential application of LCO 3.0.4.b.

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify... Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 7 days 7 days Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time.

Performance of the Surveillance initiates the subsequent 7 days interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an extension of the time interval to 1.25 times the interval specified in the Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to facility be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the facility interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Examples 1.4-3 and 1.4-4), then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR (continued)

Dresden 2 and 3 1.4-2 Amendment No. 212/204

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued) is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.

EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2.

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

(continued)

Dresden 2 and 3 1.4-3 Amendment No. 185/180

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

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

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 25% RTP.

Perform channel adjustment. 7 days The interval continues whether or not the unit operation is

< 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (plus the extension allowed by SR 3.0.2) with power 25% RTP.

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

(continued)

Dresden 2 and 3 1.4-4 Amendment No. 232/225

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-4 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

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

Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.

Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2),

but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would require satisfying the SR.

Dresden 2 and 3 1.4-5 Amendment No. 185/180

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, and LCO 3.0.8.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

a. MODE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and

b. MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1, 2, and 3.

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of (continued)

Dresden 2 and 3 3.0-1 Amendment No. 259/252

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 the results, determination of the acceptability of (continued) entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.11, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

(continued)

Dresden 2 and 3 3.0-2 Amendment No. 255/248

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Special Operations LCOs is optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations LCO shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.

LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:

a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; or

b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At the end of the specified period, the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met. spent fuel storage pool LCO 3.0.9 LCOs, including associated ACTIONS, shall apply to each unit individually, unless otherwise indicated. Whenever the LCO refers to a system or component that is shared by both units, the ACTIONS will apply to both units simultaneously.

Dresden 2 and 3 3.0-3 Amendment No. 259/252

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Dresden 2 and 3 3.0-4 Amendment No. 255/248

SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

SR 3.0.5 SRs shall apply to each unit individually, unless otherwise indicated.

spent fuel storage pool Dresden 2 and 3 3.0-5 Amendment No. 255/248

Spent Fuel Storage Pool Water Level 3.7.8 FACILITY 3.7 PLANT SYSTEMS 3.7.8 Spent Fuel Storage Pool Water Level LCO 3.7.8 The spent fuel storage pool water level shall be 19 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool, During movement of new fuel assemblies in the spent fuel storage pool with irradiated fuel assemblies seated in the spent fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage A.1 --------NOTE---------

pool water level not LCO 3.0.3 is not within limit. applicable.

Suspend movement of Immediately fuel assemblies in the spent fuel storage pool.

7 days SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify the spent fuel storage pool water In accordance level is 19 ft over the top of irradiated with the fuel assemblies seated in the spent fuel Surveillance storage pool racks. Frequency Control Program Dresden 2 and 3 3.7.8-1 Amendment No. 237/230

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location Boundary 4.1.1 Site and Exclusion Area Boundaries The site area boundary follows the Illinois River to the north, the Kankakee River to the east, a country road from Divine extended eastward to the Kankakee River on the south, and the Elgin, Joliet, and Eastern Railway right-of-way on the west. The exclusion area boundary shall be an 800 meter radius from the centerline of the reactor vessels.

4.1.2 Low Population Zone Deleted The low population zone shall be a five mile radius from the centerline of the reactor vessels.

4.2 Reactor Core Deleted 4.2.1 Fuel Assemblies The reactor shall contain 724 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. The assemblies may contain water rods or other assembly bypass channels. Limited substitutions of Zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide and hafnium metal as approved by the NRC.

(continued)

Dresden 2 and 3 4.0-1 Amendment No. 220/211

TS Markups unchanged from Attachment 2 of Dresden Administrative Controls LAR submitted Responsibility on September 24, 2020 (ML20269A404) 5.1 5.0 ADMINISTRATIVE CONTROLS plant facility 5.1 Responsibility 5.1.1 The station manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. shift shift manager 5.1.2 A unit supervisor shall be responsible for the control room command function (Since the control room is common to both units, the control room command function for both units can be satisfied by a single unit supervisor). During any absence of the unit supervisor from the control room while the unit is in MODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the unit supervisor from the control room while the unit is in MODE 4 or 5 or defueled, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

Dresden 2 and 3 5.1-1 Amendment No. 185/180

TS Markups unchanged from Attachment 2 of Organization Dresden Administrative Controls LAR submitted 5.2 on September 24, 2020 (ML20269A404) 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations facility staff Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

the safe storage facility and handling of a. Lines of authority, responsibility, and communication shall irradiated fuel be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional Exelon descriptions of departmental responsibilities and relationships, and job descriptions for key personnel Decommissioning positions, or in equivalent forms of documentation. These requirements, including the generic titles of those personnel fulfilling the responsibilities of the positions plant delineated in these Technical Specifications, shall be documented in the Quality Assurance Manual. Program facility

b. The station manager shall be responsible for overall safe storage handling operation of the plant and shall have control over those onsite activities necessary for safe operation and responsible maintenance of the plant.

irradiated fuel

c. A corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures the safe storage needed to ensure acceptable performance of the staff in and handling of operating, maintaining, and providing technical support to irradiated fuel the plant to ensure nuclear safety.

CERTIFIED FUEL HANDLERS

d. The individuals who train the operating staff, or perform facility to ensure radiation protection or quality assurance functions, may safe storage and report to the appropriate onsite manager; however, these handling of individuals shall have sufficient organizational freedom to irradiated fuel ensure their independence from operating pressures.

Facility ability to perform their assigned functions 5.2.2 Unit Staff The unit staff organization shall include the following:

(continued) facility Dresden 2 and 3 5.2-1 Amendment No. 225/217

Each duty shift shall be composed of at least one shift manager and two Organization NON-CERTIFIED OPERATORS. The NON-CERTIFIED OPERATOR 5.2 position may be filled by a CERTIFIED FUEL HANDLER.

5.2 Organization TS Markups unchanged from Attachment 2 of Dresden Administrative Controls LAR submitted on September 24, 2020 (ML20269A404) 5.2.2 Unit Staff (continued)

Facility a. A total of three non-licensed l operators for the two units is required in all conditions. At least one of the required non-licensed operators shall be assigned to each unit.

during the b. Shift crew composition may be less than the minimum movement of requirement of 10 CFR 50.54(m)(2)(i) and Specifications irradiated fuel and 5.2.2.a and 5.2.2.f for a period of time not to exceed 2 during movement hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to of loads over restore the shift crew composition to within the minimum irradiated fuel requirements.

c. A radiation protection technician shall be on site when fuel Oversight of is in the reactor. The position may be vacant for not more irradiated fuel than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required handling operations position.

shall be provided by a CERTIFIED FUEL d. Deleted.

HANDLER.

e. The T operations manager or shift operations supervisor shall The shift manager shall hold an SRO license.

be a CERTIFIED FUEL HANDLER. f. The Shift Technical Advisor (STA) shall provide advisory technical support to the shift manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

At all times when irradiated fuel is stored in the spent fuel pools, at least one person qualified to and the following conditions are met:

stand watch in the control room (NON-CERTIFIED OPERATOR or CERTIFIED FUEL 1) No movement of irradiated fuel is in progress; HANDLER) shall be present in the control room. 2) No movement of loads over irradiated fuel is in progress.

This provision does not permit any shift crew position to be unstaffed upon shift change due to the absence or tardiness of an oncoming shift crew member Dresden 2 and 3 5.2-2 Amendment No. 231/224

TS Markups unchanged from Attachment 2 of Dresden Administrative Unit Staff Qualifications Facility Controls LAR submitted on September 5.3 Facility 24, 2020 (ML20269A404) 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications facility 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the Exelon Quality Assurance Topical Report.

5.3.2 An NRC-approved training and retraining program for Decommissioning CERTIFIED FUEL Quality Assurance HANDLERS shall Program be maintained.

Dresden 2 and 3 5.3-1 Amendment No. 258/251

TS Markups unchanged from Attachment 2 of Dresden Administrative Controls LAR submitted Procedures on September 24, 2020 (ML20269A404) 5.4 5.0 ADMINISTRATIVE CONTROLS applicable to the safe storage and 5.4 Procedures handling of irradiated fuel 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; Deleted

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, Section 7.1;

c. Fire Protection Program implementation; and

d. All programs specified in Specification 5.5.

Dresden 2 and 3 5.4-1 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) (continued) shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primary Coolant Sources Outside Containment Deleted This program provides controls to 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. The systems include the Core Spray, High Pressure Coolant Injection, Low Pressure Coolant Injection, Isolation Condenser, Shutdown Cooling, Reactor Water Cleanup, process sampling (until such time as a modification eliminates the PASS penetration as a potential leakage path), containment monitoring, and Standby Gas Treatment. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at 24 month intervals.

The provisions of SR 3.0.2 are applicable to the 24 month Frequency for performing integrated system leak test activities.

5.5.3 Deleted.

(continued)

Dresden 2 and 3 5.5-2 Amendment No. 197/190

Programs and Manuals 5.5 5.5 Programs and Manuals 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 shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

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

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

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

(continued)

Dresden 2 and 3 5.5-3 Amendment No. 230/223

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

1. For noble gases: a dose rate 500 mrems/yr to the whole body and a dose rate 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrems/yr to any organ;

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

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

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Control Program Surveillance Frequencies.

5.5.5 Component Cyclic or Transient Limit Deleted This program provides controls to track the UFSAR Section 3.9, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 DELETED (continued)

Dresden 2 and 3 5.5-4 Amendment No. 254/247

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) Deleted The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

(continued)

Dresden 2 and 3 5.5-5 Amendment No. 254/247

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

Tests described in Specification 5.5.7.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of adsorber operation; after any structural maintenance on the charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.7.d and 5.5.7.e shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI/ASME N510-1980 at the system flowrate specified below:

ESF Ventilation System Penetration Flowrate Standby Gas < 1.0% 3600 cfm and Treatment (SGT) 4400 cfm System Control Room < 0.05% 1800 scfm and Emergency 2200 scfm Ventilation (CREV)

System

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI/ASME N510-1980 at the system flowrate specified below:

(continued)

Dresden 2 and 3 5.5-6 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Penetration Flowrate Standby Gas < 1.0% 3600 cfm and Treatment (SGT) 4400 cfm System Control Room < 0.05% 1800 scfm and Emergency 2200 scfm Ventilation (CREV)

System

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30C and relative humidity (RH) specified below:

ESF Ventilation System Penetration RH Standby Gas Treatment 2.5% 70%

(SGT) System Control Room 0.5% 70%

Emergency Ventilation (CREV) System

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P Flowrate Standby Gas 6 inches 3600 cfm and Treatment (SGT) water guage 4400 cfm System Control Room 6 inches 1800 scfm and Emergency water guage 2200 scfm Ventilation (CREV) System (continued)

Dresden 2 and 3 5.5-7 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate that the heaters for each of the ESF systems dissipate the value, corrected for voltage variations at the 480 V bus, specified below when tested in accordance with ANSI/ASME N510-1989:

ESF Ventilation System Wattage Standby Gas Treatment (SGT) 27 kW and System 33 kW Control Room Emergency 10.8 kW and Ventilation (CREV) System 13.2 kW 5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Off-Gas System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the Off-Gas System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Waste Management System is 0.7 curies in each tank and 3.0 curies total in all tanks, which is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

(continued)

Dresden 2 and 3 5.5-8 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Diesel Fuel Oil Testing Program Deleted A diesel fuel oil testing program shall establish the required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits,

3. A clear and bright appearance with proper color or water and sediment within limits;

b. Within 31 days following addition of the new fuel oil to storage tanks verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits; and

c. Total particulate concentration of the fuel oil in the storage tanks is 10 mg/l when tested every 31 days in accordance with the applicable ASTM Standard.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

5.5.10 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or (continued)

Dresden 2 and 3 5.5-9 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals DSAR 5.5.10 Technical Specifications (TS) Bases Control Program (continued)

2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

DSAR

d. Proposed changes that meet the criterion of Specification 5.5.10.b.1 or 5.5.10.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.11 Safety Function Determination Program (SFDP) Deleted This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

a. The SFDP shall contain the following:

1. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

2. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4. Other appropriate limitations and remedial or compensatory actions.

(continued)

Dresden 2 and 3 5.5-10 Amendment No. 185/180

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

b. A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power or loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

3. A required system redundant to support system(s) for the supported systems described in b.1 and b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.12 Primary Containment Leakage Rate Testing Program Deleted

a. This program shall establish the leakage testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

(continued)

Dresden 2 and 3 5.5-11 Amendment No. 257/250

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

b. The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 43.9 psig.

c. The maximum allowable primary containment leakage rate, La, at Pa, is 3% of primary containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Primary containment overall leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests.

2. Air lock testing acceptance criteria is the overall air lock leakage rate is 0.05 La when tested at Pa.

e. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

5.5.13 Battery Monitoring and Maintenance Program Deleted This Program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," including the following:

a. Actions to restore battery cells with float voltage

< 2.13 V, and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

(continued)

Dresden 2 and 3 5.5-12 Amendment No. 257/250

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Control Room Envelope Habitability Program Deleted A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation (CREV) System, CRE occupants 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. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by the CREV system, operating at the flow rate required by the VFTP, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered (continued)

Dresden 2 and 3 5.5-13 Amendment No. 226/218

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Control Room Envelope Habitability Program (continued) inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.15 Surveillance Frequency Control Program Deleted This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Dresden 2 and 3 5.5-14 Amendment No. 237/230

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 (Deleted) 5.6.2 Annual Radiological Environmental Operating Report

---

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

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

facility The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (continued)

Dresden 2 and 3 5.6-1 Amendment No. 214/206

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

(ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Report

---

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

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year in facility accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

facility 5.6.4 (Deleted) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) Deleted

a. 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 for the following:

1. The APLHGR for Specification 3.2.1.

2. The MCPR for Specification 3.2.2.

(continued)

Dresden 2 and 3 5.6-2 Amendment No. 214/206

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3. The LHGR for Specification 3.2.3.

4. Control Rod Block Instrumentation Setpoint for the Rod Block MonitorUpscale Function Allowable Value for Specification 3.3.2.1.

5. The OPRM setpoints for the trip function for SR 3.3.1.3.3

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."

2. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

3. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."

4. WCAP-16081-P-A, "10x10 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."

5. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 2 to Code Description, Qualification and Application."

6. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."

7. WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement 1."

8. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287."

9. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

(continued)

Dresden 2 and 3 5.6-3 Amendment No. X251/244

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

10. WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application," Revision 1, October 2011.

11. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.

12. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"

Advanced Nuclear Fuels Corporation, May 1995.

13. EMF-85-74(P) Revision 0 Supplement 1 (P)(A) and Supplement 2 (P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.

14. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.

15. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors -

Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

16. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.

17. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

18. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999.

19. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000.

(continued)

Dresden 2 and 3 5.6-4 Amendment No. 251/244

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. EMF-2209(P)(A) Revision 3, "SPCB Critical Power Correlation," AREVA NP, September 2009.

21. ANP-10298P-A Revision 1, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA, March 2014.

22. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.

23. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient ThermalHydraulic Core Analysis," Exxon Nuclear Company, February 1987.

24. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.

25. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.

26. EMF-2292 (P)(A) Revision 0, "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

27. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.

28. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

(continued)

Dresden 2 and 3 5.6-5 Amendment No. 251/244

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report Deleted When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline 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 OPERABLE status.

Dresden 2 and 3 5.6-6 Amendment No. X251/244

Attachment 4 Markup of Technical Specifications Bases Pages (For Information Only)

Dresden Nuclear Power Station Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 MARKED UP UNITS 2 AND 3 TECHNICAL SPECIFICATIONS BASES PAGES B 3.0-1 through B 3.0-23 B 3.7.8-1 through B 3.7.8-2

LCO Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY and LCO 3.0.2 BASES LCOs LCO 3.0.1 through LCO 3.0.9 establish the general 3.7 requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated.

facility LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered, unless otherwise specified.

The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the (continued)

Dresden 2 and 3 B 3.0-1 Revision 77

LCO Applicability B 3.0 BASES LCO 3.0.2 remedial measures that permit continued operation of the (continued) unit that is not further restricted by the Completion Time.

In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Condition no longer exists. The individual LCO's ACTIONS specify the Required Actions where this is the case.

An example of this is in LCO 3.4.9, "RCS Pressure and Temperature (P/T) Limits."

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The ACTIONS for not meeting a single LCO adequately manage any increase in plant risk, provided any unusual external conditions (e.g., severe weather, offsite power instability) are considered. In addition, the increased risk associated with simultaneous removal of multiple structures, systems, trains or components from service is assessed and managed in accordance with 10 CFR 50.65(a)(4). Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) are entered.

(continued)

Dresden 2 and 3 B 3.0-2 Revision 78

LCO Applicability B 3.0 BASES (continued)

LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:

Deleted.

a. An associated Required Action and Completion Time is not met and no other Condition applies; or

b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.

This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. Planned entry into LCO 3.0.3 should be avoided. If it is not practicable to avoid planned entry into LCO 3.0.3, plant risk should be assessed and managed in accordance with 10 CFR 50.65(a)(4), and the planned entry into LCO 3.0.3 should have less effect on plant safety than other practicable alternatives.

Upon entering LCO 3.0.3, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to enter lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.

(continued)

Dresden 2 and 3 B 3.0-3 Revision 78

LCO Applicability B 3.0 BASES LCO 3.0.3 A unit shutdown required in accordance with LCO 3.0.3 may be (continued) terminated and LCO 3.0.3 exited if any of the following occurs:

a. The LCO is now met.

b. The LCO is no longer applicable.

c. A Condition exists for which the Required Actions have now been performed.

d. ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for entering the next lower MODE applies. If a lower MODE is entered in less time than allowed, however, the total allowable time to enter MODE 4, or other applicable MODE, is not reduced. For example, if MODE 3 is entered in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, then the time allowed for entering MODE 4 is the next 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />, because the total time for entering MODE 4 is not reduced from the allowable limit of 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to enter a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in (continued)

Dresden 2 and 3 B 3.0-4 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.3 LCO 3.7.8, "Spent Fuel Storage Pool Water Level." LCO 3.7.8 (continued) has an Applicability of "During movement of irradiated fuel assemblies in the spent fuel storage pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.8 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.8 of "Suspend movement of fuel assemblies in the spent fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3.

These exceptions are addressed in the individual Specifications.

LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other Deleted. specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with either LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered following entry into the MODE or other specified condition in the Applicability will permit continued operation within the MODE or other specified condition for an unlimited period of time.

Compliance with ACTIONS that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made and the Required Actions followed after entry into the Applicability.

For example, LCO 3.0.4.a may be used when the Required Action to be entered states that an inoperable instrument channel must be placed in the trip condition within the Completion Time. Transition into a MODE or other specified condition in the Applicability may be made in accordance with LCO 3.0.4 and the channel is subsequently placed in the (continued)

Dresden 2 and 3 B 3.0-5 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.4 tripped condition within the Completion Time, which begins (continued) when the Applicability is entered. If the instrument channel cannot be placed in the tripped condition and the subsequent default ACTION ("Required Action and associated Completion Time not met") allows the OPERABLE train to be placed in operation, use of LCO 3.0.4.a is acceptable because the subsequent ACTIONS to be entered following entry into the MODE include ACTIONS (place the OPERABLE train in operation) that permit safe plant operation for an unlimited period of time in the MODE or other specified condition to be entered.

LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed (continued)

Dresden 2 and 3 B 3.0-6 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.4 MODE change is acceptable. Consideration should also be (continued) given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability.

LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.

LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically (continued)

Dresden 2 and 3 B 3.0-7 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.4 applied to Specifications which describe values and (continued) parameters (e.g., RCS Specific Activity), and may be applied to other Specifications based on NRC plant-specific approval.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4.

Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

(continued)

Dresden 2 and 3 B 3.0-8 Revision 72

LCO Applicability B 3.0 BASES (continued)

LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been Deleted. removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service; or

b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY.

This Specification does not provide time to perform any other preventive or corrective maintenance. LCO 3.0.5 should not be used in lieu of other practicable alternatives that comply with Required Actions and that do not require changing the MODE or other specified conditions in the Applicability in order to demonstrate equipment is OPERABLE.

LCO 3.0.5 is not intended to be used repeatedly.

An example of demonstrating equipment is OPERABLE with the Required Actions not met is opening a manual valve that was closed to comply with Required Actions to isolate a flowpath with excessive Reactor Coolant System (RCS) Primary Containment Isolation Valve (PCIV) leakage in order to perform testing to demonstrate that RCS PCIV leakage is now within limit.

Examples of demonstrating equipment OPERABILITY include instances in which it is necessary to take an inoperable channel or trip system out of a tripped condition that was directed by a Required Action, if there is no Required Action Note for this purpose. An example of verifying OPERABILITY of equipment removed from service is taking a tripped channel out of the tripped condition to permit the logic to function and indicate the appropriate response during performance of required testing on the inoperable channel.

(continued)

Dresden 2 and 3 B 3.0-9 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.5 Examples of demonstrating the OPERABILITY of other equipment (continued) are taking an inoperable channel or trip system out of the tripped condition 1) to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system, or 2) to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.

The administrative controls in LCO 3.0.5 apply in all cases to systems or components in Chapter 3 of the Technical Specifications, as long as the testing could not be conducted while complying with the Required Actions. This includes the realignment or repositioning of redundant or alternate equipment or trains previously manipulated to comply with ACTIONS, as well as equipment removed from service or declared inoperable to comply with ACTIONS.

LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Deleted. Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system's LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the plant is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.

When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability.

However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCO's Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the plant is maintained in a safe condition in the support system's Required Actions.

(continued)

Dresden 2 and 3 B 3.0-10 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.6 However, there are instances where a support system's (continued) Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Specification 5.5.11, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.

Cross division checks to identify a loss of safety function for those support systems that support safety systems are required. The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.

If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

This loss of safety function does not require the assumption of additional single failures or loss of offsite power.

Since operation is being restricted in accordance with the ACTIONS of the support system, any resulting temporary loss of redundancy or single failure protection is taken into account. Similarly, the ACTIONS for inoperable offsite circuit(s) and inoperable diesel generator(s) provide the necessary restriction for cross division inoperabilities.

This explicit cross division verification for inoperable AC electrical power sources also acknowledges that supported system(s) are not declared inoperable solely as a result of inoperability of a normal or emergency electrical power (continued)

Dresden 2 and 3 B 3.0-11 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.6 source (refer to the definition of OPERABLEOPERABILITY).

(continued)

When a loss of safety function is determined to exist, and the SFDP requires entry into the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists, consideration must be given to the specific type of function affected. Where a loss of function is solely due to a single Technical Specification support system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction source due to low tank level) the appropriate LCO is the LCO for the support system. The ACTIONS for a support system LCO adequately addresses the inoperabilities of that system without reliance on entering its supported system LCO. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system.

LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit.

These special tests and operations are necessary to Deleted.

demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions. Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect.

The Applicability of a Special Operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Special Operations LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be followed. When a Special Operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCO's (continued)

Dresden 2 and 3 B 3.0-12 Revision 72

LCO Applicability B 3.0 BASES LCO 3.0.7 Applicability (i.e., should the requirements of this other (continued) LCO not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO). However, there are instances where the Special Operations LCO's ACTIONS may direct the other LCOs' ACTIONS be met. The Surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO. If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to be met concurrent with the requirements of the Special Operations LCO.

LCO 3.0.8 LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended Deleted. safety function when associated snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS) under licensee control. LCO 3.0.8 applies to snubbers that only have seismic function.

It does not apply to snubbers that also have design functions to mitigate steam/water hammer or other transient loads. The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii), and, as such, are appropriate for control by the licensee.

If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported system's LCO(s) must be declared not met and the conditions and Required Actions entered in accordance with LCO 3.0.2.

LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. LCO 3.0.8.a allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the snubber(s) before declaring the supported system inoperable.

(continued)

Dresden 2 and 3 B 3.0-13 Revision 77

LCO Applicability B 3.0 BASES LCO 3.0.8 The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable based on the low (continued) probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system.

LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to restore the snubber(s) before declaring the supported system inoperable. The 12-hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function.

The following configuration restrictions shall be applied to the use of LCO 3.0.8:

a. At least one high pressure makeup path (e.g., using High Pressure Coolant Injection (HPCI) or Isolation Condenser (IC) or its equivalent) and heat removal capability (e.g., suppression pool cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s),

OR

b. At least one low pressure makeup path (e.g., Low Pressure Coolant Injection (LPCI) or Core Spray (CS)) and heat removal capability (e.g., suppression pool cooling or shutdown cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s).

(2) LCO 3.0.8.b can only be used following verification that at least one success path exists, using equipment not associated with the inoperable snubber(s), to provide makeup and core cooling needed to mitigate Loss of Offsite Power (LOOP) accident sequences (i.e., initiated by a seismically-induced LOOP event (continued)

Dresden 2 and 3 B 3.0-14 Revision 77

LCO Applicability B 3.0 BASES LCO 3.0.8 with concurrent loss of all safety system trains (continued) supported by the out-of-service snubbers).

Each use of LCO 3.0.8 requires confirmation that at least one train (or subsystem) of systems supported by the inoperable snubbers would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads. LCO 3.0.8 does not apply to non-seismic snubbers. In addition, a record of the design function of the inoperable snubber (i.e., seismic vs. non-seismic), implementation and compliance with the configuration restrictions defined above, and the associated plant configuration shall be available on a recoverable basis for NRC inspection.

LCO 3.0.8 requires that risk be assessed and managed.

Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (i.e., the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.

LCO 3.0.8 does not apply to non-seismic functions of snubbers. The provisions of LCO 3.0.8 apply to seismic snubbers that may also have non-seismic functions provided the supported systems would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads. Non-seismic snubber issues will be addressed in the corrective action program.

LCO 3.0.9 LCO 3.0.9 establishes the applicability of each Specification to both Unit 2 and Unit 3 operation. Whenever a requirement applies to only one unit, or is different for each unit, this will be identified in the appropriate section of the Specification (e.g., Applicability, Surveillance, etc.) with parenthetical reference, Notes, or other appropriate presentation within the body of the requirement.

spent fuel storage pools Dresden 2 and 3 B 3.0-15 Revision 77

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES 5 SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated.

SR 3.0.2 and SR 3.0.3 apply in Chapter 5 only when invoked by a Chapter 5 Specification.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

met LCOs Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this LCOs Specification, however, is to be construed as implying that systems or components are OPERABLE when:

are met when a. The systems or components are known to be inoperable, the although still meeting the SRs; or facility

b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR.

Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment (continued)

Dresden 2 and 3 B 3.0-16 Revision 77

SR Applicability B 3.0 BASES SR 3.0.1 because the ACTIONS define the remedial measures that apply.

(continued) Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process are:

a. Control Rod Drive maintenance during refueling that requires scram testing at 800 psig. However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.4.3 is satisfied, the control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psig to perform other necessary testing.

b. High pressure coolant injection (HPCI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPCI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."

interval.

(continued)

Dresden 2 and 3 B 3.0-17 Revision 77

SR Applicability B 3.0 facility BASES SR 3.0.2 SR 3.0.2 permits a 25% extension of the interval specified (continued) in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g.,

transient conditions or other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. Therefore, when a test interval is specified in the regulations, the test interval cannot be extended by the TS, and the SR includes a Note in the Frequency stating "SR 3.0.2 is not applicable."

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per..." basis. The 25%

extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%

extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

(continued)

Dresden 2 and 3 B 3.0-18 Revision 77

SR Applicability B 3.0 BASES (continued)

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not that an been performed within the specified Frequency. A delay LCO is period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified not met Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met. This delay period provides adequate time to perform Surveillances that have been missed. This delay period permits the performance of a Surveillance before complying with Required Actions or other remedial measures that might preclude performance of the Surveillance.

The basis for this delay period includes consideration of facility unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered not to have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

(continued)

Dresden 2 and 3 B 3.0-19 Revision 77

SR Applicability B 3.0 BASES SR 3.0.3 SR 3.0.3 is only applicable if there is a reasonable (continued) expectation the associated equipment is OPERABLE or that variables are within limits, and it is expected that the Surveillance will be met when performed. Many factors should be considered, such as the period of time since the Surveillance was last performed, or whether the Surveillance, or a portion thereof, has ever been performed, and any other indications, tests, or activities that might support the expectation that the Surveillance will be met when performed. An example of the use of SR 3.0.3 would be a relay contact that was not tested as required in accordance with a particular SR, but previous successful performances of the SR included the relay contact; the adjacent, physically connected relay contacts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not sufficient to infer the behavior of the associated equipment from the performance of similar equipment. The rigor of determining whether there is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance. If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will be met when performed. For Surveillances that have not been performed for a long period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERABLE. The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used repeatedly to extend Surveillance intervals. While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration (continued) facility Dresden 2 and 3 B 3.0-20 Revision 77

SR Applicability B 3.0 BASES SR 3.0.3 changes required or shutting the plant down to perform the (continued) Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182,

'Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants.' This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the licensee's Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

(continued)

Dresden 2 and 3 B 3.0-21 Revision 77

SR Applicability B 3.0 BASES (continued)

SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

variable limits This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the the safety of the Applicability for which these systems and components ensure facility safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the variables to failure to exercise the good practice of restoring systems within limits or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with an LCO LCO 3.0.4.

is not However, in certain circumstances, failing to meet an SR met will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that variables outside surveillances do not have to be performed on inoperable specified limits equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.

SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.

(continued)

Dresden 2 and 3 B 3.0-22 Revision 77

SR Applicability B 3.0 BASES SR 3.0.4 The provisions of SR 3.0.4 shall not prevent entry into (continued) MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

SR 3.0.5 SR 3.0.5 establishes the applicability of each Surveillance to both Unit 2 and Unit 3 operation. Whenever a requirement applies to only one unit, or is different for each unit, this will be identified with parenthetical reference, Notes, or other appropriate presentation within the SR.

spent fuel storage pools Dresden 2 and 3 B 3.0-23 Revision 77

Spent Fuel Storage Pool Water Level B 3.7.8 FACILITY B 3.7 PLANT SYSTEMS B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

A general description of the spent fuel storage pool design is found in the UFSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident are found in Reference 2.

APPLICABLE The water level above the irradiated fuel assemblies is an SAFETY ANALYSES explicit assumption of the fuel handling accident. A fuel handling accident is evaluated to ensure that the radiological consequences (calculated control room operator dose and doses at the exclusion area and low population zone boundaries) are below the 10 CFR 50.67 (Ref. 3) exposure guidelines. A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as discussed in Regulatory Guide 1.183 (Ref. 4).

The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core. The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.

The spent fuel storage pool water level satisfies in the spent fuel Criterion 2 of 10 CFR 50.36(c)(2)(ii). storage pool LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.

(continued)

Dresden 2 and 3 B 3.7.8-1 Revision 31

Spent Fuel Storage Pool Water Level B 3.7.8 BASES (continued)

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool or whenever movement of new fuel assemblies occurs in the spent fuel storage pool with irradiated fuel assemblies seated in the spent fuel storage pool, since the potential for a release of fission products exists.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of fuel assemblies is not a sufficient reason to require a reactor shutdown.

When the initial conditions for an accident cannot be met, action must be taken to preclude the accident from occurring. If the spent fuel storage pool level is less than required, the movement of fuel assemblies in the spent fuel storage pool is suspended immediately. Suspension of this activity shall not preclude completion of movement of a fuel assembly to a safe position. This effectively precludes a spent fuel handling accident from occurring.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued) 7 days. The 7 day Frequency is acceptable, based on operating experience, considering that the water volume in the pool is normally stable, and all water level changes are controlled by facility procedures.

Dresden 2 and 3 B 3.7.8-2 Revision 55