RS-21-036, Response to Request for Additional Information Re Dresden License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

Response to Request for Additional Information Re Dresden License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

ML21075A340
Person / Time
Site:	Dresden
Issue date:	03/16/2021
From:	Simpson P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-21-036
Download: ML21075A340 (27)

Text

4300 Winfield Road Warrenville, IL 60555 Exelon Generation 630 65 7 2000 Office RS-21-036 10 CFR 50.90 March 16, 2021 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:

Response to Request for Additional Information (RAI) Regarding Dresden, Units 2 and 3, License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

References:

1. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "License Amendment Request-Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition," dated October 29, 2020 (NRC Accession No. ML20303A313)

2. Email from Russell Haskell (U.S. Nuclear Regulatory Commission) to Mitchel Mathews (Exelon Generation Company, LLC), "RAls for Dresden Nuclear Power Station, Units 2 and 3, re: Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition ," dated March 2, 2021 (NRC Accession No. ML21063A320)

By application dated October 29, 2020 (Reference 1), Exelon Generation Company, LLC (Exelon) submitted a License Amendment Request to revise the Dresden Nuclear Power Station, Units 2 and 3 (Dresden), Renewed Facility Operating Licenses (RFOLs) and the associated Technical Specifications (TS) consistent with the permanent cessation of reactor operations and permanent defueling of the reactors. The revised RFOLs and TS will be identified as the Dresden Permanently Defueled Technical Specifications (POTS).

In an email dated March 2, 2021 (Reference 2), the U.S. Nuclear Regulatory Commission (NRC) identified areas where additional information is needed to complete the review of Reference 1. The draft RAI questions in Reference 2 were discussed during a teleconference between Exelon and NRC representatives on February 11, 2021. The NRC has requested that EGC provide a response to the questions posed in Reference 2 by April 1, 2021 . The Attachments to this letter provide the requested information .

March 16, 2021 U.S. Nuclear Regulatory Commission Page 2 EGC has reviewed the information supporting the No Significant Hazards Consideration and the Environmental Consideration that was previously provided to the NRC in Attachment 1 of the Reference 1 letter. The additional information provided in this submittal does not affect the conclusion that the proposed license amendment does not involve a significant hazards consideration. This additional information also does not affect the conclusion that there is no need for an environmental assessment to be prepared in support of the proposed amendment.

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

Exelon is providing a copy of this letter and its attachments to the State of Illinois.

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 16th day of March 2021 .

Respectfully, Patrick R. Simpson Sr. Manager - Licensing Exelon Generation Company, LLC Attachments: 1. Response to NRC Request for Additional Information

2. Markup of Permanently Defueled Technical Specifications License Amendment Request Attachment 1 Pages

3. Revised Markup of Technical Specifications Pages cc: w/ Attachments NRC Regional Administrator, Region Ill 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 Response to NRC Request for Additional Information In an email dated March 2, 2021 (Reference 1), the U.S. Nuclear Regulatory Commission provided the following request for additional information related to the October 29, 2020, license amendment request (LAR) (Reference 2) from Exelon Generation Company, LLC (Exelon) regarding proposed defueled Technical Specifications and revised License Conditions for permanently defueled conditions at Dresden Nuclear Power Station, Units 2 and 3 (Dresden) as follows:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 REQUEST FOR ADDITIONAL INFORMATION RE:

PROPOSED DEFUELED TECHNICAL SPECIF/CA T/ONS AND REVISED LICENSE CONDITIONS FOR PERMANENTLY DEFUELED CONDITION DOCKET NOS. 50-237 AND 50-249 By application dated October 29, 2020 (Agencywide Documents Access and Management System Accession No. ML20303A313), Exelon Generation Company, LLC, requested a License Amendment Request re: "Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition" for Dresden Nuclear Power Station, Units 2 and 3. The proposed changes would revise the renewed facility operating licenses and TSs consistent with the permanent cessation of operations and permanent defueling of the reactors.

The provisions in 10 CFR 50.36(c)(6), "Decommissioning," in part, apply only to nuclear power reactor facilities that have submitted the certifications required by 10 CFR 50.82(a)(1). For such facilities, technical specifications 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. In the application the licensee states, "[t]his 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."

To complete its review of the TS changes, the U.S. Nuclear Regulatory Commission (NRC) staff requests the following additional information:

Discussion: fRAl-1a & 1b) - re: Final SafetyAnalvsis Report fFSARJ The Final Safety Analysis Report (FSAR) is the principal document upon which the NRC bases its safety evaluation supporting the issuance of an operating license for a nuclear power plant. The updated FSAR (UFSAR) incorporates changes made to the FSAR in accordance with 10 CFR 50.71(e). The UFSAR serves as a major source of information on the current plant design and supporting analyses.

Page 1 of 11

Attachment 1 Response to NRC Request for Additional Information NRC decommissioning guidance (e.g., RG 1.184) discusses that the FSAR, which provides a licensing basis for the evaluation of licensing activities under 10 CFR 50.59, will have to be updated to cover decommissioning activities.

The Dresden license amendment request (LAR) Attachment 1 states, "[t]he Technical Specifications Bases Control Program is being modified to reflect that once the facility is permanently defueled the title of the UFSAR will be revised to DSAR." LAR Attachment 3, "Markup of Technical Specifications Pages," reflects this proposed change in a markup of TS 5.5.10, "The Technical Specifications (TS) Bases Control Program," by replacing UFSAR with DSAR. DSAR is not a term that is described, defined, or required by NRC regulation. The proposed change, UFSAR revised to DSAR, occurs in two instances within Dresden TS 5.5.10 (emphasis added in bold italic):

Dresden TS 5.5.10.b currently states:

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

2. a change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

Dresden TS 5.5.10.c currently states:

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

Staff Requests:

Given that NRC regulations, such as 10 CFR 50.59, are written in terms of FSAR, and DSAR is not a term that is described, defined, or required in NRC regulations; (RAl-1a): Please explain how the DSAR (replaces UFSAR in TS 5.5.10.b.2 above) will remain subject to the provisions of 10 CFR 50.59?

Exelon Response to RAl-1 a:

The Defueled Safety Analysis Report (DSAR) is the UFSAR retitled to reflect a permanently shutdown and defueled facility. The same regulations and controls that currently apply to the existing UFSAR (e.g., 10 CFR 50.59 and 10 CFR 50.71(e)) will continue to apply to the DSAR.

By proposing to replace the document title of UFSAR with DSAR, there was no underlying intention of relaxing or modifying any applicable requirements.

Page 2 of 11

Attachment 1 Response to NRC Request for Additional Information Additionally, 10 CFR 50.71(e)(6) states, "The updated FSAR {UFSAR] shall be maintained by the licensee until the Commission terminates their license." Given NRC requirements for licensees to maintain the updated FSAR until the Commission terminates their license and that a DSAR is not described, defined, or required in NRC regulations.

(RAl-1b): Please explain how the Dresden TS Bases (DSAR replaces UFSAR in TS 5.5.10.c above) will be maintained consistent with the updated FSAR under this proposed title change?

Exelon Response to RAl-1 b:

As stated in the response to RAl-1 a above, the same regulations and controls that currently apply to the existing UFSAR (e.g., 10 CFR 50.59) will continue to apply to the DSAR.

Therefore, the Dresden TS Bases will be maintained consistent with the DSAR as it was maintained consistent with the updated FSAR. A revised markup of the TS page that defines the DSAR acronym is provided in Attachment 3.

Discussion: fRAl-2) - Regulatory Basis/Issue re: New Fuel Handling Accident Analvsis Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000 provides the methodology for analyzing the radiological consequences of several design-basis accidents (DBAs) to show compliance with 10 CFR 50.67. Regulatory Guide 1.183 provides guidance to licensees on acceptable application of alternate source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," (SRP) Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Rev. 0, July 2000 provides review guidance to the staff for the review of alternative source term amendment requests. Section 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in RG 1.183. The dose acceptance criteria for the fuel handling accident (FHA) are a Total Effective Dose Equivalent (TEDE) of 6.3 rem at the exclusion area boundary (EAB) for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the control room for the duration of the accident.

In Attachment 1 of the LAR, Exelon states that a new Fuel Handling Accident (FHA) analysis was performed to determine the Control Room (CR), Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses at Dresden. This analysis is required to replace the current FHA which only covers FHA events due to a fuel assembly being dropped on the reactor core, which currently bounds the radiological consequences of an FHA in the Spent Fuel Pool.

Page 3 of 11

Attachment 1 Response to NRC Request for Additional Information Staff Request:

The LAR indicates that the analysis conforms to RG 1.183 and that the limits of RG 1.183 continue to be met, however, many of the assumptions and parameters used are not specified and the doses calculated are not specified.

(RAl-2): Please provide sufficient technical details of the new FHA, covering FHAs in the spent fuel pool post-cessation of power operations, to allow for evaluation. The NRC staff requests this information include but not necessarily be limited to; computer program(s) used to calculate dose, key input variables, calculated dose to CR, EAB, and LPZ, source term use in the new FHA analysis, fall height of fuel assembly, water coverage, decontamination factors, and atmospheric dispersion factors.

Exelon Response to RAl-2:

Exelon intends to implement a new FHA analysis at Dresden following cessation of power operations. The new FHA analysis was performed to determine the CR, EAB, and LPZ doses at Dresden due to an FHA in the Reactor Building post-cessation of power operations, after all fuel has been removed from the reactors. This analysis conforms to the Regulatory Guide 1.183 (Reference 3) and Regulatory Issue Summary (RIS) 2006-04 (Reference 4) methodology and provides limiting doses for Dresden. The FHA within the Drywell was determined to no longer be applicable with the reactors permanently defueled.

The analysis assumes that the activity in the Reactor Building is released through the Reactor Building Vent Exhaust Stack with no credit taken for accident mitigation by the Control Room Emergency Ventilation System or the Standby Gas Treatment system. This FHA analysis concludes that a decay time of 48 days is required to meet the limits of 10 CFR 50.67 (Reference 5) and Regulatory Guide 1.183 (Reference 3).

The following provides additional technical information related to the FHA analysis for post-cessation of power operations provided in Attachment 1 of Reference 1. This information does not supersede the decommissioning FHA information in Attachment 1 of Reference 1.

Some of the information is repeated from Attachment 1 of Reference 1 for completeness in summarizing the FHA analysis.

Computer Programs Used to Calculate Dose In accordance with Regulatory Guide (RG) 1.183, Section 3.1 (Reference 3), ORIGEN-ARP was used to calculate the core inventory at various decay times post-cessation of power operations.

RADTRAD Version 3.03 was used to calculate the CR, EAB, and LPZ doses following an FHA in the Reactor Building at various decay times post-shutdown. The doses to the CR, EAB, and LPZ calculated using RADTRAD are provided later in this response.

Page 4 of 11

Attachment 1 Response to NRC Request for Additional Information Key Input Variables The design inputs for the new FHA analysis are provided in Table 1 below.

T a bl e 1 Des1gn

. nputs f or New FHA A na1ys1s I .

Parameter Value Assigned Discussion/Reference

=

2957*1.02 3,016.14 Rated Thermal Power 3,016 MWt Renewed Facility Operating Licenses Isotopic Core Inventory (Ci/MWt) Table 2 below N/A Fraction of Fission Product Inventory in Gap Group Fraction Regulatory Guide 1.183, 1-131 0.08 "Alternative Radiological Source Kr-85 0.10 Terms for Evaluating Design Other Noble Gases 0.05 Basis Accidents at Nuclear Other Haloqens 0.05 Power Reactors," Section 3.2, Alkali Metals 0.12 Table 3 (Reference 3)

Radionuclide Composition Group Elements Noble Gases Xe, Kr Regulatory Guide 1.183, Section Haloqens I, Br 3.4, Table 5 Alkali Metals Cs , Rb Dresden, Units 2 and 3 Updated Radial Peaking Factor 1.7 Final Safety Analysis Report (UFSAR) Section 15.7.3.4.2.4.2 Dresden, Units 2 and 3 Updated 116 (W Optima2)

Damaged Fuel Rods Final Safety Analysis Report, 179 (A ATRIUM10)

Section 15.7.3.4.1 Number of Fuel Assemblies in Core 724 UFSAR, Table 1.2-1 SFP Water Depth <:: 19 feet UFSAR, Section 15.7.3.4.2.4.2 Reactor Buildinq (RB) Volume 4 .50E+06 ft 3 UFSAR, Section 6.2.3.2 Flow Rate from RB to Environment 300,000 ft 3/min Ref. 9 Decontamination Factors {OF) of Iodine Overall Effective DF, Total Iodine 135 Ref. 9 Chemical Form of Iodine Released from Pool Water. Calculated based on decontamination factors and form released to SFP Elemental 80%

Ref. 9 Organic 20%

Other DF of Noble Gases 1 Regulatory Guide 1.183, App. B Duration of Release 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Regulatory Guide 1.183, App. B Control Room Aerosol Elemental 0% Not credited in the analysis Organic CR Volume 64,000 ft 3 Assumption in analysis Page 5 of 11

Attachment 1 Response to NRC Request for Additional Information T a bl e 1 Des1gn

. nputs f or New FHA A na1ys1s I .

Parameter Value Assigned Discussion/Reference Unfiltered Intake Flow (0-isolation) 64,000 ft 3/min Assumption in analysis Unfiltered lnleakage (isolation-30 days) 10 ft3/min Ref. 8 CR Charcoal Filter Efficiency 0%

Not credited in the analysis CR HEPA Filter Efficiency 0%

CR Occupancy Factors (OF)

Time (hours) OF(%)

0-24 100 Regulatory Guide 1.183, Position 24-96 60 4.2.6 96- 720 40 Regulatory Guide 1.183, Section CR Breathing Rate 3.5E-04 m 3/sec 4.1.3 CR Atmospheric Dispersion Factors Time (hours) xtQ (sec/m 3) 0-2 6.44E-04 2-8 4.91 E-04 Ref. 7 8 -24 2.02E-04 24 - 96 1.36E-04 96 - 720 1.05E-04 LPZ Atmospheric Dispersion Factors Time (hour) xtQ (sec/m 3 )

0-2 2.63E-05 2-8 1.09E-05 Ref. 7 8-24 7.02E-06 24-96 2.70E-06 96- 720 6.86E-07 Site Area Boundary (EAB) Parameters Regulatory Guide 1.183, Section Breathing Rate 3.5E-04 m 3/sec 4.1.3 0-2 hr Atmospheric Dispersion Factor 2.51 E-04 sec/m 3 Ref. 7, Table 4-1 (X/Q)

Offsite Breathing Rates (LPZ)

Breathing Rate Time (hours) Regulatory Guide 1.183, Section (m 3/sec) 4 .1.3 (Conservative assumption 0-8 3.5E-04 in analysis that breathing rate 8-24 3.5E-04 continues at 3.5E-04 m3 /sec) 24- 720 3.5E-04 Source Term Used in the New FHA Analysis ORIGEN-ARP was used to calculate the core inventory at various decay times post-cessation of power operations. Irradiation cases were performed within ORIGEN-ARP, modeling full power operation for 1,714.6 Effective Full Power Days (EFPD). The EFPDs used, as determined by the product of maximum core weight (130 Metric Tons of Uranium (MTU)) and maximum burnup (39,000 MWD/MTU), and rated thermal power (2,957 MWt), bounds any potential operating Page 6 of 11

Attachment 1 Response to NRC Request for Additional Information history, as any nominal operating cycle will incur natural reductions in energy production via maintenance evolutions and down powers. The fraction of the core fuel damaged is based on the current UFSAR design basis of a postulated rupture of the cladding of all the fuel rods in the dropped assembly. 1-131 was increased by a factor of 1.6 to account for additional fractional release relative to other iodine isotopes. Kr-85 was increased by a factor of 2.0 to account for additional fractional release relative to other Noble Gas isotopes. The nuclide inventory, without Kr-85 and 1-131 adjustments applied, in the FHA at 48 days after shutdown is as follows:

Table 2: Source Term (Noble Gas and Iodine) at 48 days post-shutdown Decommissioning FHA 48 Day Decay Isotope Specific Activity (Ci/MWth)

Kr-85 5.10E+02 Kr-85m O.OOE+OO Kr-87 O.OOE+OO Kr-88 O.OOE+OO 1-131 4.51E+02 1-132 1.46E+OO 1-133 1.21E-12 1-134 O.OOE+OO 1-135 O.OOE+OO Xe-133 1.16E+02 Xe-135 1.77E-33 Calculated Dose to CR EAB. and LPZ The acceptance criteria for the EAB and LPZ doses are based on 10 CFR 50 .67 (Reference 5) and Table 6 of RG 1.183 (Reference 3):

1. 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, should not receive a radiation dose in excess of 6.3 rem TEDE (Total Effective Dose Equivalent).

2. An individual located at any point on the outer boundary of the LPZ, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), should not receive a radiation dose in excess of 6.3 rem TEDE.

One objective of the new FHA analysis is to determine the decay time required such that post-FHA doses remain below the regulatory limits listed above without credit for Reactor Building Ventilation or isolation or Control Room Emergency Ventilation. RADTRAD cases were run assuming 40, 45,47, 48, 49, 55, 65, 75, 100, 150, 300, and 365 days of decay. Based on a sensitivity study, it is assumed that the control room is isolated after three (3) minutes.

Page 7 of 11

Attachment 1 Response to NRC Request for Additional Information The minimum decay time required, without credit for Reactor Building Ventilation or isolation or Control Room Emergency Ventilation to meet 10 CFR 50.67 CR limits and RG 1.183, Table 6 EAB and LPZ limits was determined . Based on the results in Table 3, all applicable regulatory limits can be met after 48 days of decay.

Table 3: Post-Cessation of Power Operation CR, EAB, and LPZ Doses Decay Time (rem TEDE)

(days)

CR Site Boundary (EAB) LPZ 40 9.7271E+OO 2.3425E-02 2.4548E-03 45 6.3219E+OO 1.5205E-02 1.5934E-03 47 5.3214E+OO 1.2797E-02 1.3411 E-03 48 4.8817E+OO 1.1741 E-02 1.2304E-03 49 4.4786E+OO 1.0773E-02 1.1289E-03 55 2.6697E+OO 6.4385E-03 6.7472E-04 65 1.1280E+OO 2.7588E-03 2.8911 E-04 75 4.7742E-01 1.2115E-03 1.2696E-04 100 5.6736E-02 2.1347E-04 2.2371 E-05 150 2.3400E-03 8.4079E-05 8.8111 E-06 300 1.5580E-03 8.0161E-05 8.4005E-06 365 1.5407E-03 7.9272E-05 8.3073E-06 Regulatory Limit 5.00 6.30 6.30 Fall Height of Fuel Assembly The assumed fall height of the assembly is maintained at 34 feet, which is the height used in the current/operating Fuel Handling Accident analysis and represents a conservative assumption .

Water Coverage and Decontamination Factors It is expected that all fuel will be in the Spent Fuel Pool post-cessation of operations, so the FHA is modeled assuming 19 feet of water coverage. The associated overall effective iodine Decontamination Factor (DF) for 19 feet of water coverage is calculated to be 135, with gap activity fractions consistent with Regulatory Guide 1.183.

Atmospheric Dispersion Factors The atmospheric dispersion factors (x/Qs) used in this analysis are from Reference 7. Table 1 provides the specific atmospheric dispersion factors for the new FHA analysis.

Page 8 of 11

Attachment 1 Response to NRC Request for Additional Information In addition to the above RAls, the NRC staff has identified the following apparent editorial errors:

1. There appears to be an editorial error in Attachment 1 of the LAR (pg. 75185). In the detailed description of the proposed changes to Dresden TS Section 5.5.10, "Technical Specifications (TS) Bases Control Program," there is a reference to the use of the acronym "DSAR" for the first time in the LAR without spelling it out or providing a description/definition. If DSAR is retained (could be changed based on a response to other questions), please spell out the acronym on it first use in the LAR. This lack of a defined termed is similarly reflected in the LAR Attachment 3, TS 5.5.10 TS markup page (document pg. 1651201).

2. There appears to be an editorial error in Attachment 1 of the LAR (pg. 40185). In the detailed description of the proposed changes to Dresden TS Section 1.1, "Definitions," for CERTIFIED FUEL HANDLER, the column entitled "Basis for Change" states, " ... the LAR proposing changes to TS Sections 1.1and5.0 (Reference 2) which is currently under NRC review." However, the citing of (Reference 2) appears to be incorrect because (Reference 2) does not propose any changes to TS Sections 1.1or5.0. Please identify the correct reference and update the LAR, as appropriate.

3. There is a conflict between LAR Attachment 1 (description of changes) and LAR Attachment 3 (TS markups) regarding proposed changes to TS Section 1.3, "Completion Times." There should be no difference between the changes proposed in LAR Attachment 1 and those reflected in LAR Attachment

3. However, LAR Attachment 3 deletes two additional words " ... ensuring safe ... ".

Please identify the desired proposed change and update the LAR as appropriate.

4. There appears to be an editorial error in Attachment 1 of the LAR in the basis discussion for proposed changes to LCO 3.0.1 (pg. 48185). The last sentence in the LCO 3.0.1 basis discussion refers to deleting the reference to LCO 3.0.9. Since LCO 3.0.1 currently does not refer to LCO 3.0.9, this appears to be an editorial error. Please identify the correct reference and update the LAR as appropriate.

5. There appears to be an editorial error in Attachment 1 of the LAR in the basis discussion (summary section)(pg. 57185) for proposed changes to TS Section 3.3, "Instrumentation." In the bottom half of the summary section, there is a reference to a proposed new license condition 2.EE (Dresden Unit 3). However, in LAR Attachment 1, in the section which describes proposed changes to license conditions (pg.38185), refers to the proposed new license condition as 3.EE. Please identify the correct number for the proposed new license condition and update the LAR as appropriate (note, LAR Attachment 1 refers to 2.EE several times and is not limited to TS Section 3.3).

6. There appears to be an editorial error in Attachment 1 of the LAR in the basis discussion for proposed changes to TS 5.5.14. The last sentence in the TS 5.5.14 basis discussion refers to TS 5.5.13. Based on the context, this appears to be an Page 9 of 11

Attachment 1 Response to NRC Request for Additional Information editorial error. Please identify the correct TS reference and update the LAR as appropriate.

Exelon Response:

A markup of the affected pages of the October 29, 2020, submittal that addresses the errors identified by the NRC in Items 1-6 above is provided in Attachment 2. Revised markups of the Technical Specifications pages that are affected by the errors identified above are provided in .

References:

1. Email from Russell Haskell (U.S. Nuclear Regulatory Commission) to Mitchel Mathews (Exelon Generation Company, LLC), "RAls for Dresden Nuclear Power Station, Units 2 and 3, re: Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition ," dated March 2, 2021

2. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition,"

dated October 29, 2020 (NRC Accession No. ML20303A313)

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

4. Regulation Issue Summary (RIS) 2006-04, " Experience with Implementation of Alternative Source Terms," dated March 7, 2004

5. 10 CFR 50.67, as published in 64 FR 72001, dated December 3, 1999

6. Calculation DRE04-0030, Revision 2, "Atmospheric Dispersion Factors (X/Qs) for Accident Release," (previously provided in ADAMS Accession No. ML061460305 as Attachment 1)

7. Exelon Calculation DRE04-0030, Revision 2, "Atmospheric Dispersion Factors (X/Qs) for Accident Release," Previously provided as supporting information for Dresden, Units 2 and 3 Request for License Amendment Related to Application of Alternative Source Term (ADAMS Accession No. ML052520217)

8. NUREG-0800, "Section 6.4, Control Room Habitability System," Revision 3, dated March 2007

9. Calculation DRE02-0036, "Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms," Revision 1, Previously provided as supporting information for Dresden, Units 2 and 3 Request for License Amendment Related to Application of Alternative Source Term (ADAMS Accession No. ML052430326)

Page 10 of 11

Attachment 1 Response to NRC Request for Additional Information

10. Letter from Maitri Banerjee (U.S. NRC) to Christopher M. Crane (Exelon Generation Company), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Re: Adoption of Alternative Source Term Methodology (TAC Nos. MB6530, MB6531 , MB6532, MB6533, MC8275, MC8276, MC8277 and MC8278)," dated September 11, 2006 (ADAMS Accession No. ML062070292)

Page 11 of 11

Attachment 2 Markup of Permanently Defueled Technical Specifications License Amendment Request Attachment 1 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 ATTACHMENT 1 PAGES 2 of 85 40 of 85 48 of 85 57 of 85 63 of 85 64 of 85 65 of 85 68 of 85 75 of 85 76 of 85

Attachment 1 Evaluation of Proposed Changes Related Licensing Actions ~

By letter dated September 24, 2020 (Reference ~). 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 POTS 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 OBA 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 OBA 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 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 POTS. The details and justification for the proposed changes are provided, arranged by TS section.

TS Section 1.0 - Use and Application rTS Being Deleted rrs Being Retained 1.1 - Definitions 1.2 - Lo~ical 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 lonqer aooly to the permanently defueled facility status .

Definition Basis for Change AVERAGE PLANAR LINEAR HEAT This definition is proposed for deletion in the POTS since the term is GENERATION RATE (APLHG?:fl- not used in any POTS specification . This term is only meaningful to

~eactor authorized to operate.

CERTIFIED FUEL HANDLER A :~r Certified Fuel Handler was proposed for addition to the P in the LAR proposing changes to TS Sections 1.1 and 5.0 (Referenc which is currently under NRC review .

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

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

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

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

DOSE EQUIVALENT 1-131 This definition is proposed for deletion in the POTS since the term is not used in any POTS 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 1-131 is used for dose analysis of accidents involving primary coolant releases . Those accident conditions will no lonqer aooly to the permanently shutdown and defueled facility.

DRAIN TIME This defin ition is proposed for deletion in the POTS since the term is not used in any POTS specification. This term is no longer aoolicable since fuel will be permanently removed from the reactors .

Page 40 of 85

Attachment 1 Evaluation of Proposed Changes completion of the Required Action(s) is not required, ceA=l~letion ef the ReEj1::1irel Actien(s) is net reEjblirel ,

unless otherwise stated . blnless etherwise statel .

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 l:ffif.t spent fuel storage pool individually, Whenever the LCO refers to a system or component blnless etherwise inlicatel. Whenever the bCO that is shared by both units, the ACTIONS will apply ref.ers te a systeA=l er ceA=l~enent that is sharel ey to both units simultaneously. eeth blnits, the ACTIO~J~ will a~~ly te eoth blnits 13.0.81 Basis LCO 3.0.1 is modified by elimina1 ng the references to MODES because this term does not apply to a facility in the permanently defuele d condition . MODES as defined in Table 1.1-1 are defined for operating or refueling conditions . Table 1 . ~ 71 was proposed for deletion in TS Section 1.0. In addition , the references to LCOs 3.0.7 and ~ 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 POTS. 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 POTS 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 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 3.EE The above TS are related t assuring the appropriate functional capability of plant equipment, and control of process variables, desig features , or operating restrictions required for safe operation of the facility only when the reactor is in ODES 1 through 5, during movement of recently irradiated fuel in the secondary containment, or hen DGs are required during Modes 4 and 5. After the certifications required by 10 CFR 50 .82( )(1) are submitted for Dresden, the 10 CFR 50 licenses will no longer authorize operation of the r actor or emplacement or retention of fuel in the reactor vessels pursuant to 10 CFR 50 .82(a)(2). Techn cal 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 consid red "recently irradiated," and after 48 days of decay following permanent shutdown , the need for filter d recirculation of control room air will not be required , precluding the need for the DGs to support plan quipment associated with control room ventilation. Proposed License Conditions 2.C.(22) and ~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 (SOC) System-Hot Shutdown Page 57 of 85

Attachment 1 Evaluation of Proposed Changes 3.EE Summary The above TS are related t assuring the appropriate functional capability of plant equipment, and control of process variables, desig features, or operating restrictions required for safe operation of the facility only when the reactor is in ODES 1 through 3 or during movement of recently irradiated fuel in the secondary containment. A er the certifications required by 10 CFR 50.82(a)(1) are submitted for Dresden , the 10 CFR 50 lie nses will no longer authorize operation of the reactor or emplacement or retention of fuel in the react r 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 nalysis there are no active systems credited as part of the initial conditions of the analysis or as part of th primary success path for mitigation of the FHA with the unit permanently defueled. Therefore, the u e of these systems is not credited or required in the FHA for reduction of nuclides or a reduction of o site 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 da after permanent shutdown through the imposition of proposed License Conditions 2.C.(22) and . 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 Conditioninq (AC) System 3.7.6 - Main Condenser Offqas 3.7.7 - Main Turbine Bypass System 3.7.8 - Spent Fuel Storaqe 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 POTS with the chanqes 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 OBA or transient.

The CCSW System is operated whenever the containment cool ing 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.

3.EE Te n cification 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System ,"

pro ides the operability requirements for the Control Room Emergency Ventilation AC system. The fun tion of the Control Room Emergency Ventilation AC system is to provide temperature control for the con ol room emergency zone following isolation of the control room emergency zone. The Control Room Em rgency Ventilation AC System is designed to provide a controlled environment under both normal and ccident conditions. This specification is applicable during Modes 1, 2 and 3, or during movement of rec ly irradiated fuel assemblies in the secondary containment. Proposed License Conditions 2.C.(22) and ~ 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 onl when the reactor is in MODES 1 throu h 3 or durin movement of recentl irradiated fuel in the Page 64 of 85

Attachment 1 Evaluation of Proposed Changes 13.EEI secondary containment. After the certifications re quired by 10 CFR 50.82(a)(1) are submitted for Dresden , the 10 CFR 50 licenses will no longer a ~ thorize operation of the reactor or emplacement or retention of fuel in the reactor vessels pursuant tc 10 CFR 50.82(a)(2). As discussed in the Fuel Handling Accident Analysis for the Permanently Defueled C ondition section of this attachment, in the post permanent shutdown FHA analysis there are no c: ctive systems credited as part of the initial conditions of the analysis or as part of the primary success pat n 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 of site doses after 48 days of decay time. Exelon proposes to prohibit movement of irradiated fuel c: fter the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 d ~ ts after permanent shutdown through the imposition of the proposed License Conditions 2.C .(22) and ~ 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 POTS 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 POTS 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 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 13.EEI The above TS are related to assuring the appropriate functic nal capability of plant equipment, and control of process variables , design features , or operating restrictior s required for safe operation of the facility only when the reactor is in MODES 1 through 5, during mov1 ment 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 ! O licenses will no longer authorize operation of the reactor or emplacement or retention of fuel in the reac or 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 analys s there are no active systems credited as part of the initial conditions of the analysis or as part of the pr mary success path for mitigation of the FHA with the units permanently shutdown and defueled. Therefon , 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 irradiat ~d fuel after the submittal of the certification of permanent removal of fuel from the reactor vessel until 48 lltiys after permanent shutdown through the imposition of the proposed License Conditions 2.C.(22) and ~ 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 IDefueled Safety Analysis Report (DSAR) h TS Section 5.5.9 - Diesel The Diesel Fuel Oil Testi 1g Program was established to implement the Fuel Oil Testing Program required testing of both n ~w and stored fuel oil for the EDGs. This program is proposed for E imination from the POTS since the EDGs will not perform any safety fur ction in the permanently shutdown and defueled facility. The TS< ssociated with the EDGs and the diesel fuel oil subsystem (TS 3.8.1, 3.B.2, and 3.8.3) are proposed for removal from the POTS as describe j above.

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

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 POTS 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 aooly 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(0) and Testing Program 10 CFR 50 Appendix J, Option B, as modified by exemptions. This program will not be retained in the POTS, 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 13.EE ~ a smoke challenge. Following permanent shutdown and defueling, and

~ ~ through the imposition of the proposed License Conditions 2.C.(22) and

- * ~ 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 15.5.14 I following the establ shment of the permanent shutdown and defueled condition. Addition 3lly, as previously discussed, TS 3.3.7.1 and 3.7.4 are not proposed t~,.t>e included in the POTS ; thus, Technical Specification ~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 POTS 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 POTS 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 POTS 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 POTS, 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 aooly 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 POTS 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 3 Revised 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 REVISED MARK UPS OF UNITS 2 AND 3 TECHNICAL SPECIFICATIONS PAGES 1.3-1 5.5-10

Completion Times

1. 3 1.0 USE AND APPLICATION storage and handling of 1.3 Completion Times irradiated fuel I

PURPOSE The purpose of this section \ s to establish the Completion Time convention and to provi ~ guidance for its use.

BACKGROUND Limiting Conditions for Opera :\ on (LCOs) specify minimum requirements for ensuring safe operation of the 1:1nit . The ACTIONS associated with an LCO state Conditions that typica l ly describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated 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., inoperab l e eq1:1ipment or variable not within limits) that requires entering an

~' ACTIONS Condition unless otherwise specified, providing the

~ tl-R-l-t is in a MO Q~ or specified condition stated in the Applicability of the LCO. ldnless otFler',1ise specifieEI , tFle Comp l etion Ti me beg i ns .: hen a sen i or licensed operator on 1

the operating sh i ft ere'.: ',1ith responsibil i ty for plant operations makes the determination that an LCO is not met and an /\CTIO~l £ Condition is entered . The "othen1ise specified " exceptions are varied , s1:1ch as a Req1:1ired Action Note or £1:1rveillance Req1:1irement Note that provides an alternative time to perform specific tasks , s1:1ch as testing ,

11itho1:1t start i ng the Completion Time . Whi l e 1:1tilizing the Note , sho1:1ld a Condition be app l icab l e for any reason not addressed by the Note , the Comp l etion Time begins . £ho1:1ld the time allo1i'ance in the ~l ote be e><ceeded , the Completion Time begins at that point . The exceptions may also be incorporated into the Completion Time . For e><ample , LCO 2 . g . 1 , "AC £01:1rces Operating ," Req1:1ired Action B. 2 ,

req1:1ires declaring req1:1ired feat1:1re(s) s1:1pported by an inoperable di es el generator , inoperable li'hen the red1:1ndant req1:1ired feat1:1re(s) are inoperable . The Completion Time states , " ~ ho1:1rs from discovery of Condi ti on g conrnrrent 1

dith inoperability of red1:1ndant req1:1ired feat1:1re(s) ." -+/--i+

this case the Comp l etion Time does not begin 1:1nti l 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. 255/2~g

Programs and Manuals 5.5 5.5 Programs and Manuals DefueledSafetyAnalysisReport(DSAR) 5.5.10 Technical S ecifications (continued)

2. A change to the 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 ~ .

d. Proposed changes that meet the criterion of Specifica ~

5.5.10.b.l 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.?l(e).

5. 5 .11  :;Z£:!:!: a:t:fe~tE:'t=.'::ff:::+/-b!:+/-JInllic+/-t+/-i~ottn::::JiJi:le~t:d:ettr:!!:m!::+/-i:t+/-n:!:!: a:t:t+/-:!iOst:!n+/-:::::!P::tr:l2o~g+/-r:!!:a!+/-+/-m:::::b(£2::!F"::!:Ot:tP: z): .<?-------ll

. De Ieted I This program enswres loss of safety function is detected and appropriate actions taken . Upon entry into LCD J . 0 . § , an evaluation shal l 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 reswlt of the swpport system inoperability and corresponding exception to entering swpported system Condition and ReEJwired Actions . +-1+-i--&

program implements the reEjwirements of LCD J . 0. § .

.a-. The HIJP shall contain the follo1.'ing :

h Provisions for cross division checks to enswre a loss of the capability to perform the safety f~nction ass~med in the accident analysis does not go wndetected; b Provisions for enswring the plant is maintained in a safe condition if a loss of fwnction condition exists; J-. Provisions to enswre that an inoperable s~pported system ' s Completion Time is not inappropriately extended as a reswlt of mwltiple swpport system inoperabilities;

~

4-.- Other appropriate limitations and remedial or compensatory actions .

(continued)

Dresden 2 and 3 5.5-10 Amendment No. HHi/lgQ