ML20265A240

Issuance of Amendment Nos. 272 and 265 to Increase Allowable Main Steam Isolation Valve Leakage

ML20265A240
Person / Time
Site:	Dresden
Issue date:	10/23/2020
From:	Haskell R Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co
Haskell R
References
EPID L-2019-LLA-0232
Download: ML20265A240 (57)

Text

October 23, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENT NOS. 272 AND 265 TO INCREASE ALLOWABLE MAIN STEAM ISOLATION VALVE LEAKAGE (EPID L-2019-LLA-0232)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 272 to Renewed Facility Operating License No. DPR-19 and Amendment No. 265 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3 (DNPS), respectively. The amendments are in response to your application dated October 21, 2019, as supplemented by letters dated May 6, 2020, and August 24, 2020.

The amendments revise both the individual and the combined main steam isolation valve (MSIV) leakage rate limit for the four main steam lines in Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement (SR) 3.6.1.3.10; add a new TS 3.6.2.6, Drywell Spray; and revise TS 3.6.4.1, Secondary Containment, SR 3.6.4.1.1. Implementation of the amendments will also include revision of the Updated Final Safety Analysis Report, as described in your letter dated May 6, 2020.

Additionally, the application included a request for a permanent exemption from certain requirements in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, Option B,Section III.A, and Section III.B, to permit exclusion of MSIV leakage from the overall integrated leak rate Type A test measurement and to permit exclusion of the MSIV pathway leakage contributions from the combined leakage rate of all penetrations and valves subject to Type B and Type C tests, respectively. By letter dated October 5, 2020 (Agencywide Document Access Management System (ADAMS) Accession No. ML20248H565), the NRC granted the licensee an exemption from these 10 CFR Part 50, Appendix J requirements, as proposed in the application.

B. Hanson A copy of the related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions [monthly] Federal Register notice.

Sincerely,

/RA/

Russell S. Haskell, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249

Enclosures:

1. Amendment No. 272 to DPR-19

2. Amendment No. 265 to DPR-25

3. Safety Evaluation cc: Listserv

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 272 License No. DPR-19

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated October 21, 2019, as supplemented by letters dated May 6, 2020, and August 24, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-19 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 272, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

Enclosure 1

3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensees letter dated May 6, 2020.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Robert F. Robert F. Kuntz Date: 2020.10.23 Kuntz 13:21:31 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-19 and Technical Specifications Date of Issuance: October 23, 2020

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 265 License No. DPR-25

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated October 21, 2019, as supplemented by letters dated May 6, 2020, and August 24, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B. of Renewed Facility Operating License No. DPR-25 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 265, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

Enclosure 2

3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensees letter dated May 6, 2020.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Robert F. Robert F. Kuntz Date: 2020.10.23 Kuntz 13:22:00 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-25 and Technical Specifications Date of Issuance: October 23, 2020

ATTACHMENT TO LICENSE AMENDMENT NOS. 272 AND 265 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating areas of change.

REMOVE INSERT License DPR-19 License DPR-19 Page 3 Page 3 License DPR-25 License DPR-25 Page 4 Page 4 Technical Specifications Replace the following pages of the Appendix A Technical Specifications with the attached revised/new pages. The revised/new pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 3.6.1.3-8 3.6.1.3-8

--- 3.6.2.6-1

--- 3.6.2.6-2 3.6.4.1-2 3.6.4.1-2

(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 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.

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 power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 272, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

Renewed License No. DPR-19 Amendment No. 272

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:

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 The Technical Specifications contained in Appendix A, as revised through Amendment No. 265, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

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

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

Renewed License No. DPR-25 Amendment No. 265

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.7 Verify each automatic PCIV actuates to In accordance the isolation position on an actual or with the simulated isolation signal. Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor In accordance instrumentation line EFCVs actuate to the with the isolation position on an actual or Surveillance simulated instrument line break signal. Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from In accordance each shear isolation valve of the TIP with the System. Surveillance Frequency Control Program SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is 62.4 scfh for Unit 2 with the and 78 scfh for Unit 3 when tested at Primary 25 psig, and the combined leakage rate Containment for all MSIV leakage paths is 156 scfh Leakage Rate for Unit 2 and 218 scfh for Unit 3 when Testing Program tested at 25 psig.

Dresden 2 and 3 3.6.1.3-8 Amendment No. 272/265

Drywell Spray 3.6.2.6 3.6 CONTAINMENT SYSTEMS 3.6.2.6 Drywell Spray LCO 3.6.2.6 Two drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One drywell spray A.1 Restore drywell spray 7 days subsystem inoperable. subsystem to OPERABLE status.

B. Two drywell spray B.1 Restore one drywell 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystems inoperable. spray subsystem to OPERABLE status.

C. Required Action and C.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 /> associated Completion Time not met. AND C.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 /> Dresden 2 and 3 3.6.2.6-1 Amendment No. 272/265

Drywell Spray 3.6.2.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify each drywell spray subsystem In accordance manual and power operated valve in the with the flow path that is not locked, sealed, or Surveillance otherwise secured in position, is in the Frequency correct position or can be aligned to the Control Program correct position.

SR 3.6.2.6.2 Verify each drywell spray nozzle is In accordance unobstructed. with the Surveillance Frequency Control Program SR 3.6.2.6.3 Verify drywell spray subsystem locations In accordance susceptible to gas accumulation are with the sufficiently filled with water. Surveillance Frequency Control Program Dresden 2 and 3 3.6.2.6-2 Amendment No. 272/265

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 ----------------NOTE---------------- In accordance Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if with the analysis demonstrates one standby gas Surveillance treatment (SGT) subsystem is capable of Frequency establishing the required secondary Control Program containment vacuum.

Verify secondary containment vacuum is 0.25 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access In accordance door in each access opening is closed, with the except when the access opening is being Surveillance used for entry and exit. Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be In accordance maintained 0.25 inch of vacuum water with the gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem Surveillance at a flow rate 4000 cfm. Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment In accordance equipment hatches are closed and sealed. with the Surveillance Frequency Control Program Dresden 2 and 3 3.6.4.1-2 Amendment No. 272/26546

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 272 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 265 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25 EXELON GENERATION COMPANY, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249

1.0 INTRODUCTION

By application dated October 21, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19294A304), as supplemented by documents dated May 6, 2020 (ADAMS Package Accession No. ML20127H890), and letter dated August 24, 2020 (ADAMS Accession No. ML20237F317), Exelon Generation Company, LLC (the licensee or Exelon), requested changes to the technical specifications (TSs) for Dresden Nuclear Power Station, Units 2 and 3 (DNPS). The proposed changes would revise both the individual and combined main steam isolation valve (MSIV) leakage rate limits for all four steam lines in TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement (SR) 3.6.1.3.10; add a new TS 3.6.2.6, Drywell Spray, to manage post-accident fission product removal; and revise TS 3.6.4.1, Secondary Containment, SR 3.6.4.1.1 to specify a modified secondary containment pressure condition.

The supplemental documents dated May 6, 2020, and August 24, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 7, 2020 (85 FR 19511).

The application also included a request for a permanent exemption from certain requirements in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, Option B,Section III.A, in order to permit exclusion of MSIV leakage from the overall integrated leak rate Type A test measurement, and Option B,Section III.B, in order to permit exclusion of the MSIV pathway leakage contributions from the combined leakage rate of all penetrations and valves subject to Type B and Type C tests. The NRC approved the exemption request by letter dated October 5, 2020 (ADAMS Accession No. ML20248H565). This is discussed below in Section 2.2.4.

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 System Descriptions 2.1.1 MSIVs The four main steam lines (MSLs), which penetrate the drywell, are automatically isolated by the MSIVs. There are two MSIVs on each steam line, one inside containment (inboard MSIV) and one outside containment (outboard MSIV). In total, these eight MSIVs are functionally part of the primary containment boundary. Leakage through these valves provides a potential leakage path for fission products to bypass secondary containment and enter the environment as a ground level release.

2.1.2 Secondary Containment The secondary containment is a structure that encloses the primary containment including components that may contain primary system fluid. The safety function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a design-basis accident (DBA) to ensure the control room (CR) operator and offsite doses are within the regulatory limits. There is no redundant train or system that can perform the secondary containment function should the secondary containment be inoperable.

The secondary containment boundary is the combination of walls, floor, roof, ducting, doors, hatches, penetrations, and equipment, that physically form the secondary containment.

Secondary containment operability is based on its ability to contain, dilute, and hold up fission products, that may leak from the primary containment following a DBA. To prevent ground level exfiltration of radioactive material while allowing the secondary containment to be designed as a mostly conventional structure, the secondary containment requires support systems to maintain the pressure at less than atmospheric pressure. During normal operation, nonsafety-related systems are used to maintain the secondary containment at a slight negative pressure to ensure any leakage is into the building and that any secondary containment atmosphere exiting the secondary containment is via a pathway monitored for radioactive material. However, during normal operation it is possible for the secondary containment vacuum to be momentarily less than is required for a number of reasons, such as during wind gusts or swapping of the normal ventilation subsystems.

2.1.3 Containment Spray Containment cooling is the operating mode of the low-pressure coolant injection (LPCI) subsystem initiated to cool the containment in the event of a loss-of-coolant accident (LOCA).

Two separate and independent containment cooling subsystems are provided to remove heat from the containment, reduce containment pressure, and restore suppression pool temperature following a postulated LOCA. Each containment cooling subsystem consists of two LPCI pumps, one containment cooling heat exchanger, one drywell spray header, and a separate spray line terminating at a common suppression chamber ring header. The containment spray mode uses LPCI pumps to deliver water from the suppression chamber through the containment cooling heat exchangers to spray headers in the drywell and/or suppression chamber. The system valve control logic permits the operator to provide any required division between containment spray flow and LPCI flow.

Containment spray cools non-condensable gases and condenses steam in the containment following a postulated LOCA. The term containment spray as used in the Updated Final Safety Analysis Report (UFSAR) refers to drywell spray and suppression chamber spray, collectively.

Drywell spray reduces drywell temperature and pressure through the combined effects of evaporative and convective cooling and is used to wash, or scrub, inorganic iodine and particulates from the drywell atmosphere into the suppression pool. The drywell spray is used during a LOCA for both the scrubbing function as well as for temperature and pressure reduction.

2.1.4 Standby Gas Treatment (SGT) and Control Room (CR) Filter Systems During emergency conditions, the SGT system is designed to be capable of drawing down the secondary containment to a required vacuum within a prescribed time and continuing to maintain the negative pressure as assumed in the accident analysis. The leak tightness of the secondary containment together with the SGT system ensure that radioactive material is either contained in the secondary containment or filtered through the SGT system filter trains before being discharged to the outside environment via the elevated release point.

Both the SGT system and the CR ventilation system have an air filtration system used to control radiation exposure during off-normal or accident conditions. The filtration systems consist of the housing that contains the filters and absorber, the filters and absorber themselves, and any interconnecting ductwork between the filter elements. The SGT system charcoal beds and high efficiency particulate air filters treat the intentional release of primary and secondary containment atmosphere to the environs in the unlikely event of a DBA and thereby reduce exposure to the public and site personnel. The CR ventilation air filtration unit provides protection from radiation exposure to allow CR access and occupancy for the duration of a LOCA with MSIV leakage at TS limits as the worst-case DBA.

2.2 Description of Proposed Changes In the license amendment request (LAR), the licensee proposed to revise both the individual and the combined main steam isolation valve (MSIV) leakage rate limit for the four MSL in TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), SR 3.6.1.3.10; add a new TS 3.6.2.6, Drywell Spray; and revise TS 3.6.4.1, Secondary Containment, SR 3.6.4.1.1.

2.2.1 TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), SR 3.6.1.3.10 establishes allowed leakage rates through each MSIV leakage path when tested at 25 pounds per square inch gauge (psig). The LAR proposes to increase this SR for each unit from 34 standard cubic feet per hour (scfh) to 62.4 scfh for Unit 2, and to 78 scfh for Unit 3, and the combined leakage rate for all MSIV leakage paths when tested at 25 psig is proposed to be increased for each unit from 86 scfh to 156 scfh for Unit 2, and to 218 scfh for Unit 3.

2.2.2 TS 3.6.2.6, Drywell Spray New TS 3.6.2.6, Drywell Spray, is proposed to be added to the DNPS TSs. Limiting condition for operation (LCO) 3.6.2.6 would require two drywell spray subsystems to be operable and has an Applicability of Modes 1, 2, and 3. LCO 3.6.2.6 would have three Actions which are:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One drywell spray A.1 Restore drywell spray 7 days subsystem inoperable. subsystem to OPERABLE status.

B. Two drywell spray B.1 Restore one drywell spray 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystems inoperable. subsystem to OPERABLE status.

C. Required Action and C.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 /> associated Completion Time not met. AND 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.2 Be in MODE 4.

Three SRs, as shown in the following table, would be associated with new LCO 3.6.2.6.

SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify each drywell spray subsystem manual and power In accordance with operated valve in the flow path that is not locked, the Surveillance sealed, or otherwise secured in position, is in the Frequency Control correct position or can be aligned to the correct position. Program SR 3.6.2.6.2 Verify each drywell spray nozzle is unobstructed. In accordance with the Surveillance Frequency Control Program SR 3.6.2.6.3 Verify drywell spray subsystem locations susceptible to In accordance with gas accumulation are sufficiently filled with water. the Surveillance Frequency Control Program 2.2.3 TS 3.6.4.1, Secondary Containment The LAR proposes to revise SR 3.6.4.1.1 by adding a Note that would allow the secondary containment vacuum limit to not be met for a short duration during normal operation, provided an analysis demonstrates that one SGT subsystem remains capable of establishing the required secondary containment vacuum. Examples of conditions when this limit may not be met include

wind gusts that lower external pressure, or loss of the normal ventilation system that maintains secondary containment vacuum. This is further discussed by the staff below in Section 3.1.1.3.

The current SR 3.6.4.1.1 states, Verify secondary containment vacuum is > 0.25 inch of vacuum water gauge.

The revised SR 3.6.4.1.1 proposes to add a Note that states, Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.

2.2.4 Exemption Request for 10 CFR Part 50, Appendix J, Option B The licensee states in the LAR that MSIV leakage will no longer be counted as part of the maximum allowable leakage rate from containment, La, aligning DNPS with the common industry practice of monitoring MSIV leakage separately from the station La (allowable leakage rate) totals. In support of this change, the licensee requested that the NRC grant an exemption from (1) the requirements of 10 CFR Part 50, Appendix J, Option B, paragraph III.A, to allow exclusion of the MSIV leakage from La measured when performing a Type A Test, and (2) the requirements of 10 CFR Part 50, Appendix J, Option B, paragraph III.B, to allow exclusion of the MSIV leakage rate of the penetration valves subject to Type B and C tests. As noted above, the NRC approved the licensees exemption request by letter dated October 5, 2020 (ADAMS Accession No. ML20248H565).

2.3 Regulatory Requirements and Guidance 2.3.1 10 CFR Part 50 The regulation at 10 CFR 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36. The applicant must include in the application, a summary statement of the bases or reasons for such specifications, other than those covering administrative controls. However, per 10 CFR 50.36(a)(1), these TS bases shall not become part of the TSs.

The regulation at 10 CFR 50.36(b) requires that each license authorizing reactor operation include TSs derived from the analyses and evaluation included in the safety analysis report and amendments thereto.

The regulation at 10 CFR 50.36(c)(2)(i) requires the TSs to include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

The regulation at 10 CFR 50.36(c)(2)(ii) requires that a TS LCO be established for each item meeting one or more of the following criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: 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.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

The regulation at 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, requires an applicant to establish a program for qualifying electric equipment that is important to safety as defined in 10 CFR 50.49(b). Subsection 50.49(e)(1) requires that the time-dependent temperature and pressure at the location of the electric equipment important to safety be established for the most severe DBA during and following which this equipment is required to remain functional. Subsection 50.49(e)(2) requires that humidity during DBAs be considered. Subsection 50.49(e)(4) requires that the radiation environment be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects. Subsection 50.49(b)(2) requires qualification of nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs 50.49(b)(1)(i)(A)-(C) by the safety-related equipment.

The regulation at 10 CFR 50.67(b)(2)(i) requires that an individual located at any point on the boundary of the exclusion area (EAB) for any 2-hour period following the onset of the postulated fission product release would not receive a radiation dose in excess of 25 roentgen equivalent man (rem) (0.25 sieverts (Sv)) total effective dose equivalent (TEDE).

The regulation at 10 CFR 50.67(b)(2)(ii) requires that an individual located at any point on the outer boundary of the low population zone (LPZ), who is exposed to the radioactive cloud resulting from the postulated fission product release would not receive a radiation dose in excess of 25 rem (0.25 SV) TEDE during the entire period of its passage.

The regulation at 10 CFR 50.67(b)(2)(iii) requires that adequate radiation protection be provided to permit access to and occupancy of the control room under accident conditions, without personnel receiving radiation exposures in excess of 5 rem (0.05 Sv) TEDE for the duration of the accident.

The regulation at 10 CFR 50.90 requires that whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to issue an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities authorized will not endanger the health and safety of the public, and will comply with the NRCs regulations.

2.3.2 General Design Criteria (GDC)

Section 3.1 of the DNPS UFSAR states that Units 2 and 3 conform with the intent of the Atomic Energy Commission GDCs for nuclear power plant construction permits. As the GDCs were finalized, the requirements were placed in Appendix A to 10 CFR Part 50. Under 10 CFR Part 50, Appendix A, GDC, Criterion 19 - Control room, a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions including LOCAs. In addition, for plants that have adopted § 50.67, with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.

2.3.3 Staff Requirements Memorandum (SRM) to SECY-19-0036 In the SRM for SECY-19-0036 (ADAMS Accession No. ML19183A408), the Commission directed the staff to apply risk-informed principles in any licensing review or other regulatory decision when strict prescriptive application of deterministic criteria is unnecessary to provide reasonable assurance of adequate protection of public health and safety. Risk-informed principles are consistent with the Commissions safety goal policy (51 FR 30028, August 21, 1986). Additionally, the NRCs policy statement on the use of Probable Risk Assessment (PRA) methods (60 FR 42622; August 16, 1995) calls for the use of PRA technology in all regulatory matters in a manner that complements the NRCs deterministic approach and supports the traditional defense-in-depth philosophy.

2.3.4 Regulatory Issue Summary (RIS)

As stated in NRC Regulatory Issue Summary (RIS) 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ADAMS Accession No. ML053460347), any licensee who chooses to reference AEB 98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Term, dated December 9, 1998 (ADAMS Accession No. ML011230531), should provide an appropriate justification that the assumptions are applicable to its particular design.

2.3.5 Regulatory Guides (RG)

The NRC staffs guidance for alternative source term (AST) reviews includes RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792).

2.3.6 Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants - NUREG-0800 The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425).

For licensees using the AST in their radiological consequence analyses, the NRC staff uses NRC regulatory guidance including NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, dated July 2000 (ADAMS Accession No. ML003734190) and NUREG-0800 (SRP) Section 6.5.2, Containment Spray as a Fission Product Cleanup System, Revision 4, dated March 2007 (ADAMS Accession No. ML070190178).

For guidance regarding human factors, the NRC staff refers to NUREG-0800, Chapter 18, Human Factors Engineering, Revision 3, dated December 2016 (ADAMS Accession No. ML16125A114).

NUREG-1764, Revision 1, Guidance for the Review of Changes to Human Actions, dated September 2007 (ADAMS Accession No. ML072640413), and NUREG-0711, Revision 3, Human Factors Engineering Program Review Model, dated November 2012 (ADAMS Accession No. ML12324A013), contains guidance on state-of-the-art human factors principles.

3.0 TECHNICAL EVALUATION

3.1 Staff Evaluation of Proposed Changes to TSs Section 2.2 above discusses the licensees proposed TS changes. The following sections (3.1.1 thru 3.1.5) include the NRC staffs evaluation of the licensees proposed TS changes associated with: (1) TSs, (2) Radiological Consequences, (3) Secondary Containment, (4) Environmental Qualifications, and (5) Credited Operator Actions (Human Factors).

3.1.1 TSs 3.1.1.1 MSIV Leakage Rate Limits (SR 3.6.1.3.10)

As discussed in Section 2.2.1 above, regarding proposed changes to SR 3.6.1.3.10, the leakage rate through each MSIV leakage path when tested at 25 psig is increased for each unit from 34 scfh to 62.4 scfh for Unit 2 and 78 scfh for Unit 3, and the combined leakage rate for all MSIV leakage paths when tested at 25 psig is increased for each unit from 86 scfh to 156 scfh for Unit 2 and 218 scfh for Unit 3.

The NRC staff has evaluated the licensees proposed MSIV leakage rate limit changes and determined they are more restrictive than the revised MSIV leakage limits used in the reanalysis

of the postulated LOCA radiological consequences, as discussed in LAR, Enclosure B (ADAMS Accession No. ML19294A306). The proposed changes to the MSIV leakage rate limits were derived from the reanalysis of the postulated LOCA radiological consequences and are more restrictive than those used in the reanalysis, and the licensee derived its proposed changes using an NRC-approved analysis. Accordingly, the staff has determined there is reasonable assurance that (1) 10 CFR 50.36(b) will continue to be met, and (2) the licensees proposed SR meets 10 CFR 50.36(c)(3), which requires that TSs include SRs for testing, calibration, and inspection, and ensures that the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met. Additionally, consistent with the regulation at 10 CFR 50.36(a)(1), the licensee included as part of the LAR the TS Bases changes that correspond to the proposed changes, and stated that these changes will be reflected in the DNPS TS Bases Control Program.

Since the licensees proposed TS changes to the current MSIV leakage rate limits meet regulatory requirements, as discussed, the NRC staff has concluded that the licensees proposed changes to SR 3.6.1.3.10 are acceptable.

3.1.1.2 Drywell Spray TS (TS 3.6.2.6)

Section 2.2.2, Drywell Spray, as proposed above, would add new TS 3.6.2.6, Drywell Spray to the DNPS TSs. The associated LCO 3.6.2.6 requires two drywell spray subsystems to be operable and has an Applicability of Modes 1, 2, and 3. Section 2.2.2 also proposes three Actions as part of TS 3.6.2.6.

The current licensing basis (CLB) as described in Chapters 6 and 15 in the DNPS UFSAR does not credit drywell spray for mitigation of any DBA or transient. However, the system and its operation are described in Chapter 6. Currently, the guidance and requirements related to the drywell spray function are maintained in the DNPS Technical Requirements Manual (TRM).

There are no existing DNPS TS requirements associated with the drywell spray function.

DNPS converted to Improved Technical Specifications (ITS), as documented in an NRC safety evaluation report supporting the issuance of license Amendments 185 and 180 (ADAMS Accession No. ML011130121).

As part of the conversion to ITS, the drywell spray requirements were removed from the DNPS TSs based on the following justification: The drywell spray function of the LPCI/containment cooling systems is utilized in post LOCA conditions to condense steam in the drywell, thereby further lowering containment pressure. Emergency operating procedures direct manual initiation of the drywell spray function of the LPCI/containment cooling systems. However, in the analysis of the bounding event for containment pressurization due to the DBA, the drywell spray function of the LPCI/containment cooling systems was not utilized for mitigation of the event.

The drywell spray function is not required for proper performance of the containment pressure suppression system and is not an initial assumption of any DBA or transient analysis.

At that time, it was determined that the requirements for the drywell spray function did not satisfy the Commissions Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132, July 22, 1993) screening criteria for remaining in the TSs, and they were relocated to the DNPS TRM, which is licensee-controlled program, in accordance with 10 CFR 50.59.

The approved amendments (i.e., Nos. 185 and 180) replaced the custom TSs and associated Bases with ITS, based on the NRC-approved guidance and criteria specified in NUREG-1433.1 10 CFR 50.36(c)(2)(ii)(C) (Criterion 3) states that a TS LCO of a nuclear reactor must be established for a structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. As stated in the Commissions Final Policy Statement, it is the intent of this criterion to capture in TSs 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 criteria), so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria.

In the LAR, the licensee submitted a reanalysis of the postulated LOCA radiological consequences. The reanalysis credits the use of drywell sprays to reduce airborne activity by scrubbing inorganic iodine and particulates from the drywell atmosphere, thereby mitigating the radiological consequences of the postulated LOCA. Because the drywell spray functions to mitigate the postulated design basis LOCA, the system meets the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii) and an LCO must be established for this system.

As discussed in Section 2.3.1 above, 10 CFR 50.36(c)(2)(i) requires TSs to include LCO(s).

The licensees proposed TS LCO 3.6.2.6 states, Two drywell spray subsystems shall be OPERABLE The licensees reanalysis of the postulated LOCA radiological consequences credits a minimum of one drywell spray subsystem to adequately scrub the inorganic iodine and particulates from the primary containment atmosphere. To ensure that these requirements are met, following a LOCA, assuming the worst-case single active component failure, two drywell spray subsystems must be operable. Therefore, the NRC staff determined that the proposed TS LCO satisfies the requirements of 10 CFR 50.36(c)(2)(i) because the LCO specifies the lowest functional capability or performance level of equipment required for safe operation of the facility.

New TS 3.6.2.6 requires two drywell spray subsystems to be operable in Modes 1, 2, and 3. In Modes 1, 2, and 3 (i.e., power operation, startup, and hot shutdown, respectively), there is considerable energy in the reactor core and a LOCA could release fission products into the primary containment. In Modes 4 and 5 (i.e., cold shutdown and refueling, respectively), there is less energy in the reactor core such that the probability and consequences of a LOCA are reduced due to the reduced pressures and temperatures in these modes. The NRC staff has determined the licensees proposed applicability is acceptable because there is less energy in the reactor core such that the probability and consequences of a LOCA are reduced due to the reduced pressures and temperatures in these modes. In addition, the applicability is consistent with the applicability of the emergency core cooling system (ECCS) TS which also functions to limit the release of radioactive materials to the environment following a LOCA.

1 Standard Technical Specifications, General Electric Plants, BWR/4, Revision 1, dated April 1995.

DNPS TSs define Actions as that part of a specification that prescribes Required Actions to be taken under designated Conditions within specified TS Completion Times.

New TS 3.6.2.6, Condition A, is entered when one drywell spray subsystem becomes inoperable. Required Action A.1 requires restoring the inoperable drywell spray subsystem to operable status within 7 days. The NRC staff concludes Action A is acceptable because the remaining operable drywell spray subsystem is adequate to perform the primary containment fission product scrubbing function and there is a low probability of a LOCA occurring during the 7-day period. The low probability of the initiating event occurrence combined with the high likelihood that the single remaining drywell spray subsystem will function make the risk associated with this condition acceptably low.

New TS 3.6.2.6, Condition B, is entered when two drywell spray subsystems become inoperable and Required Action B.1 requires restoring one drywell spray subsystem to operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The NRC staff concludes that Action B is acceptable because of the low probability of a LOCA occurring during the short TS Completion Time. The low probability of an initiating event that requires drywell spray for mitigation during this short time makes the risk associated with this condition acceptably low.

New TS 3.6.2.6, Condition C, is entered when operators are unable to restore the drywell spray subsystems to operable status within the completion time under TS Condition A or B. Required Action C.1 requires the unit to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> followed by entry to Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, as required by Required Action C.2. The NRC staff concludes that Action C is acceptable because the condition requires the operators to place the unit in a condition in which the LCO no longer applies. In addition, the proposed TS Completion Times allow a reasonable amount of time to reach Mode 3 and Mode 4 in a controlled manner and without challenging plant systems from full power operation (Mode 1).

New TS 3.6.2.6, SR 3.6.2.6.1, requires verification that each drywell spray subsystem manual and power operated valve in the flow path that is not locked sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position. New SR 3.6.2.6.2 requires verification that each drywell spray nozzle is unobstructed. New SR 3.6.2.6.3 requires verification that the drywell spray subsystem locations susceptible to gas accumulation are sufficiently filled with water.

The NRC staff concludes that the new SRs are acceptable because: (1) SR 3.6.2.6.1 ensures a flow path exists between the RHR pumps and the drywell spray nozzles so that the flow path would be available following a LOCA; (2) SR 3.6.2.6.2 ensures there are no blockages that would affect the spray pattern which would invalidate the input and assumptions used the DNPS reanalysis of the postulated LOCA radiological consequences (Enclosure B, DRE05-0048, Revision 5, of LAR); (3) SR 3.6.2.6.3 ensures the normally water-filled lines of the RHR system do not have gas accumulation. Adequate filling and venting of the piping is necessary for proper operation of the drywell spray subsystems and also prevents water hammer and pump cavitation; and, (4) existing SR 3.6.2.3.2 verifies that each required RHR pump develops a flow rate greater than or equal to 5000 gallons per minute (gpm) through the associated heat exchanger while operating in the suppression pool cooling mode, which is substantially greater than the 2,352 gpm assumed in the reanalysis of the postulated LOCA radiological consequences.

The frequency of the new SRs is in accordance with the Surveillance Frequency Control Program (SFCP). In the supplement to the LAR, the licensee stated that:

The proposed TS surveillance frequencies will be added to the SFCP at the same time the approved amendment is implemented at the site. The due dates for the existing TRM surveillances will determine the next applicable due date for SR 3.6.2.6.1 (valve position) and SR 3.6.2.6.2 (spray nozzles). The first due date for SR 3.6.2.6.3 (gas accumulation) will be set equal the next due date for SR 3.6.2.4.3 the corresponding surveillance for suppression pool spray piping.

The SR 3.6.2.6.1 frequency for valve positions will remain 31 days consistent with the current TRM surveillance. The new SR 3.6.2.6.3 checking for gas accumulation will match the corresponding suppression pool spray SR frequency of 184 days.

The SR 3.6.2.6.2 frequency for the spray nozzles will be 6 years. This frequency is shorter than the current TRM surveillance frequency of 10 years due to corrective actions resulting from a failed Unit 2 drywell header air test during the 2011 refueling outage. The test failed due to a build-up of corrosion products within the carbon steel spray header ring, which obstructed some of the nozzles and required emergent repairs. Corrective actions from 2011 included reducing the frequency of the test to 4 years for both Units 2 and 3 with an action to re-evaluate the frequency after the next surveillance. The Unit 3 surveillance was last performed in 2014 and its frequency was extended to 6 years based on historical performance (next surveillance currently scheduled for the fall 2020 outage). The Unit 2 TRM surveillance corresponding to the proposed SR 3.6.2.6.2 surveillance is currently scheduled for completion during the Unit 2 outage in fall 2019. Satisfactory results from the Unit 2 air test is expected to provide a basis for extending the frequency from 4 years to 6 years consistent with the current Unit 3 surveillance frequency. The proposed 6-year frequency is considered appropriate for the new TS SR based on the following considerations:

The ring headers are normally not wetted by the LPCI system and via the nozzles, are open to the inerted nitrogen Drywell atmosphere. Corrosion development is significantly inhibited under inerted conditions. Dresden had experienced some extended outages prior to 2011 providing opportunity for historical corrosion development, but outage length is trending shorter, providing less time for the header to be exposed to open air conditions.

The header rings on both units were cleaned to remove historical loose surface corrosion on the interior of the carbon steel spray header in 2011/2012, respectively. Subsequent inspections have not found a recurrence of the previous level of loose surface corrosion.

By procedure, the drywell spray motor operated valves 2(3)-1501-27A(B) and 2(3)-1501-28A(B) are cycled to minimize the potential for residual water from between the valves to enter the header.

Motor-operated drywell spray valves 2(3)-1501-27A(B) and 2(3)-1501-28A(B) are explicitly modeled in the Dresden Full Power Internal Events Level 1 and Level 2

PRA fault trees and the Dresden Fire PRA [probabilistic risk assessment] fault trees, capturing the drywell spray function for primary containment cooling. The drywell spray function for fission product scrubbing is implicitly modeled, where successful spray implies successful scrubbing. The drywell spray nozzles are not explicitly modeled in the PRA. However, the proposed Surveillance Requirement test intervals for Drywell Spray (31 days for SR 3.6.2.6.1, 6 years for SR 3.6.2.6.2, and 184 days for SR 3.6.2.6.3) can be represented in the Dresden PRA either explicitly or implicitly.

New TS 3.6.2.1, SR 3.6.2.6.1, SR 3.6.2.6.2, and SR 3.6.2.6.3: (1) do not reference other approved programs for the specific interval, (2) are not purely event driven, (3) are not event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs, (4) are not related to specific conditions or conditions for the performance of a SR, and (5) can be modeled either directly or implicitly in the plant-specific PRA. The NRC staff determined the base surveillance frequency interval for SR 3.6.2.6.1 is acceptable based on operating experience because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. The base surveillance frequency interval for SR 3.6.2.6.2 is acceptable because it is adequate to detect degradation in performance due to the passive nozzle design and its normally dry state and considers previous operating experience and corrective actions to monitor for corrosion.

The base surveillance frequency interval for SR 3.6.2.6.3 is acceptable because it takes into consideration the gradual nature of gas accumulation in the drywell spray system piping and the procedural controls governing system operation. With respect to SR 3.6.2.6.1, the NRC staff noted that the SR is intended to verify that each required drywell spray valve can be aligned to its required position. However, the bases for the TS state that no manipulations are required.

In the licensees response of May 6, 2020, to the staffs request for additional information regarding this matter, the licensee provided supplemental information that stated that the spray valves are normally closed motor- operated valves that are manually repositioned from the control room if required. Further, the licensee stated that the valves are tested during refueling outages in accordance with the inservice testing program to ensure that they can perform the safety function of opening following a LOCA. The NRC staff found the supplemental information adequately addressed the staffs concern.

As noted in Section 2.3.1 of this SE, 10 CFR 50.36(c)(3) requires that the TSs include SRs, which are requirements relating to test, calibration, or inspection, to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The NRC staff determined that the base surveillance frequency intervals need not be included in the TSs but can be included in the SFCP and relocated to an owner-controlled document which is administratively controlled in accordance with the TS 5.5.15, Surveillance Frequency Control Program. Therefore, the licensee continues to meet the regulatory requirements of 10 CFR 50.36(c)(3), because the necessary quality of systems and components will be maintained in accordance with the associated LCO.

In accordance with 10 CFR 50.36(a)(1), the licensee submitted TS Bases changes that correspond to the proposed TS changes for information only. The licensee stated in the LAR that it will make supporting changes to the TS Bases in accordance with TS 5.5.10, Technical Specifications (TS) Bases Control Program.

The NRC staff reviewed the proposed TS 3.6.2.6 and concludes that the TS meets the requirements of 10 CFR 50.36(a), 10 CFR 50.36(c)(2)(i), 10 CFR 50.36(c)(2)(ii), and 10 CFR 50.36(c)(3), for the reasons discussed above, and, thus, provides reasonable assurance that DNPS will have the requisite requirements and controls to operate safely.

Therefore, the NRC staff concludes that new TS 3.6.2.6, as proposed, is acceptable.

3.1.1.3 SR 3.6.4.1.1, Secondary Containment SR 3.6.4.1.1 requires the secondary containment vacuum to be 0.25 inch of vacuum water gauge when the secondary containment is required to be operable. A Note is being added to SR 3.6.4.1.1. The Note allows the SR to not be met for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if an analysis demonstrates that one SGT subsystem can establish the required secondary containment vacuum. During normal operation, conditions may occur that result in SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external pressure, or loss of the normal ventilation system that maintains secondary containment vacuum, may affect secondary containment vacuum. These conditions may not be indicative of degradations of the secondary containment boundary or of the ability of the SGT system to perform its specified safety function.

The note provides an allowance for the licensee to confirm secondary containment operability by confirming that one SGT subsystem can perform its specified safety function. This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.

While the duration of these occurrences is anticipated to be very brief, the allowance is permitted for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which is consistent with the time permitted for secondary containment to be inoperable per Condition A of LCO 3.6.4.1.

The proposed addition of the Note to SR 3.6.4.1.1 does not change the TS requirement to meet SR 3.6.4.1.3. SR 3.6.4.1.3 requires verification that the secondary containment can be maintained 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at a flow rate 4000 cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment (SGT) System, must be met; otherwise the licensee shall shut down the reactor or follow any remedial action permitted by TSs until the condition can be met.

Technical Specification Task Force (TSTF) Traveler TSTF-551, Revision 3, Revise Secondary Containment Surveillance Requirements, dated September 21, 2017, is based on the standard technical specifications (STS). The STS contain an SR (3.6.4.1.4) that the SGT be demonstrated capable of drawing the secondary containment pressure down to the required level in 120 seconds. The DNPS TSs do not contain a similar SR. However, the DNPS drawdown analysis provided in Enclosure A to the LAR demonstrates that the drawdown will occur as assumed in the AST analysis using conservative assumptions. Therefore, an SR for drawdown time is not required.

As discussed above, the DNPS secondary containment operability is based on its ability to contain, dilute, and hold up fission products that may leak from primary containment following a DBA. To prevent ground level exfiltration of radioactive material, the secondary containment pressure must be maintained at a pressure that is less than atmospheric pressure. The secondary containment requires support systems to maintain the control volume pressure less than atmospheric pressure. Following an accident, the SGT system ensures the secondary containment pressure is less than the external atmospheric pressure. During normal operation, nonsafety-related systems are used to maintain the secondary containment at a negative pressure. However, during normal operation it is possible for the secondary containment

vacuum to be momentarily less than the required vacuum for several reasons. These conditions may not be indicative of degradations of the secondary containment boundary or of the ability of the SGT system to perform its specified safety function. SR 3.6.4.1.3 requires that the SGT establish the required secondary containment pressure within 120 seconds.

LCO 3.6.4.3 requires two trains of SGT to be operable and provides required actions to restore inoperable trains or transition the plant to a condition where SGT is not required to maintain plant safety. As required by the TS, the licensees analysis demonstrates the SGTs ability to establish secondary containment pressure and maintain SGT operability. In addition, the licensees analysis confirms secondary containment operability by demonstrating that one SGT subsystem can perform its specified safety function. The NRC the staff concludes there is reasonable assurance that the secondary containment and SGT subsystem will maintain the vacuum requirements during a DBA, and the addition of this Note to SR 3.6.4.1.1, therefore, is acceptable.

Furthermore, because the specified safety functions of the secondary containment and SGT subsystem can be performed in the time assumed in the licensees accident analysis, the fission products that bypass or leak from primary containment, or that are released from the reactor coolant pressure boundary components located in the secondary containment prior to release to the environment, will be contained and processed as assumed in the licensees design basis radiological consequence dose analyses. The NRC staff, therefore, concludes that the proposed change is acceptable.

In accordance with 10 CFR 50.36(a)(1), the licensee submitted TS Bases changes that correspond to the proposed TS changes for information only. The licensee stated in the LAR that it will make supporting changes to the TS Bases in accordance with TS 5.5.10, Technical Specifications (TS) Bases Control Program.

Staff Conclusions As discussed above, the NRC staff reviewed the proposed changes to TS SR 3.6.1.3.10, new TS 3.6.2.6, and changes to TS SR 3.6.4.1.1, and determined that the requirements for TSs in 10 CFR 50.36(b) are met because the new and revised TS are derived from the analyses and ensure that plant conditions are consistent with the analyses and evaluations included in the safety analyses, including changes documented in the LAR. Additionally, the changes to the TSs were reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff concludes that the proposed TS changes meet the requirements of 10 CFR 50.36(a)(1), 10 CFR 50.36(c)(2)(i),

10 CFR 50.36(c)(2)(ii), and 10 CFR 50.36(c)(3), for the reasons discussed above, and thus provide reasonable assurance that the TSs will have the requisite requirements and controls for DNPS to operate safely. Therefore, the NRC staff concludes that the TSs, as proposed, are acceptable.

3.1.2 Radiological Consequences To demonstrate that the performance of various plant safety systems designed to mitigate the postulated radiological consequence of a design basis accident at DNPS will remain adequate after implementing the MSIV leakage rate limit increases in the requested TS changes, the LAR, Enclosure B, as supplemented included, a revision to the postulated radiological consequence analysis of the design basis LOCA. This analysis provided the results of the revised design basis LOCA radiological analysis to demonstrate compliance with 10 CFR 50.67 for the CR, EAB, and LPZ doses, and 10 CFR Part 50, Appendix A, GDC 19, for CR dose.

The revised design basis LOCA radiological analysis was performed using an NRC radiological consequence computer code, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, Version 3.03, described in NUREG/CR-6604, A Simplified Model for RADionuclide Transport and Removal And Dose Estimation (ADAMS Accession No. ML15092A284). The RADTRAD code, developed by the Sandia National Laboratories for NRC, estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The NRC staff performed independent confirmatory dose evaluations, as needed, using the RADTRAD code. The results of the evaluations performed by the licensee (Attachment 2, DRE05-0048, Revision 6, provided in Attachment 2 to the May 6, 2020 letter), as well as the applicable dose acceptance criteria from RG 1.183, Revision 0, are shown in Table 1 below.

Table 1 DNPS Units 2 and 3 Bounding LOCA Radiological Consequences Expressed as TEDE (1) (rem)

Post-LOCA Activity Release Path Post-LOCA TEDE Dose (rem)

Receptor Location CR EAB (2) LPZ (3)

Containment Leakage 2.06E-01 8.26E-02 3.23E-01 ESF Leakage 8.94E-03 5.81E-03 4.15E-02 MSIV Leakage 3.91E+00 1.98E+00 5.40E-01 Reactor Building Shine 1.77E-01 0.00E+00 0.00E+00 External Cloud Shine 5.50E-01 0.00E+00 0.00E+00 CR Filter Shine Negligible 0.00E+00 0.00E+00 Total Dose 4.86E+00 2.07E+00 9.05E-01 Acceptance Criteria 5.00E+00 2.50E+01 2.50E+01 (1) Total effective dose equivalent (2) Exclusion area boundary maximum 2-hour dose (3) Low population zone 30-day dose at the outer boundary The revised design basis LOCA radiological analysis is based, in part, on the current design basis LOCA radiological analysis approved by the NRC in License Amendment Nos. 233 and 229, Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Re: Adoption of Alternate Source Term Methodology (TAC Nos. MB6530, MB6531, MB6532, MB6533, MC8275, MC8276, MC8277 and MC8278),

dated September 11, 2006 (ADAMS Accession No. ML062070290). The amendment adopted full implementation of the AST methodology. In the NRC staffs safety evaluation (SE) issued with the amendment approving full implementation of the AST methodology, the NRC staff indicated that it had concerns regarding the use of AEB 98-03 used in the DNPS, Units 2 and 3,

current licensing basis (CLB). At that time, the NRC staff based its approval of the AST license amendment request (LAR), in part, upon additional conservatism in the MSIV leakage model, which is no longer present in the proposed LAR:

The NRC staff expressed a concern that the removal through aerosol settling was overestimated by modeling two settling volumes with the same settling velocity in each, when the settling would be expected to be at a lesser rate for the later sections of piping and at a later time considering that the larger and heavier aerosols would have already settled out of the main steam line atmosphere in upstream sections of piping. However, as stated above, Exelon

[the licensee] did not credit any reduction in drywell pressure or the MSIV leakage rate after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Leakage rates were assumed to be held constant for the entire duration of the accident for conservatism. Given this information, the NRC staff finds the Dresden and Quad Cities main steam line aerosol settling model to be reasonably conservative.

The NRC staff acknowledges that aerosol settling is expected to occur in the main steam line piping but because of recent concerns with aerosol sampling and its characteristics used in AEB-98-03 and lack of further information, the NRC staff is concerned with how much deposition (i.e., what settling velocity value) is appropriate. The licensee has used a model based on the methodology of AEB-98-03, but has applied additional conservatism (i.e. 40th percentile settling velocity, constant MSIV leakage for the entire duration of the accident) to address the NRC staffs concern about the applicability of the AEB-98-03 methodology to Dresden and Quad Cities. The NRC staff further acknowledges that the estimate of the fraction of the aerosol that leaks to the environment is uncertain because of phenomenological uncertainties concerning the environment the aerosol encounters in the various volumes assumed by Exelon.

The changes requested in this LAR modify several assumptions and inputs used to model the MSIV leakage pathway after a design basis LOCA and, thus, removed additional conservatism used by the NRC staff to find the CLB analysis reasonably conservative. For example, in the revised analysis, the RHR drywell spray system is credited for the reduction of airborne activity in the drywell by scrubbing radionuclides from the drywell air space, mitigating the consequence of the postulated design basis LOCA. The particulate iodine deposition from the RPV nozzle to the inboard MSIV, elemental iodine deposition from the RPV nozzle to the inboard MSIV, elemental iodine between inboard and outboard MSIVs, and MSIV leakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the maximum leakage are credited. To model the effect of mixing, the MSIV flow rate used in the RADTRAD model was decreased by calculating a new leak rate based on the combined volumes of the drywell and suppression chamber.

3.1.2.1 Accident Source Term/Core Inventory The licensee evaluated the offsite and CR radiological consequences due to the design basis LOCA. In accordance with RG 1.183, Revision 0, guidance, the inventory of fission products in the reactor core was determined based on the maximum full power operation of the core (3,016 megawatts thermal (MWt) This maximum full power operation is 102 percent of the 2,957 MWt-rated power. The chemical form of radioiodine released into the containment is assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide.

Except for elemental and organic iodine and noble gases, the remaining fission products are assumed to be in particulate form.

Enclosure B of the LAR provides the isotopic core inventory in Curies (Ci) for the three fuel load options: Framatome ATRIUM 10XM (Framatome), Westinghouse SVEA-96 Optima2 (SVEA),

and Global Nuclear Fuel (GNF) GNF3 fuel types that may reside in the DNPS reactor cores.

The accident source term/core inventory core average exposure (CAVEX) for the Framatome and SVEA fuel load used in the calculations is 39-gigawatt days per metric ton of uranium (U)

(GWd/MTU).

The licensee provided a new bounding core inventory using the GNF fuel load based on a combination of increased CAVEX to 43 GWd/MTU and an enrichment range between 3.7 wt percent U-235 and 4.5 wt percent U-235. This bounding core inventory covers current and anticipated future fuel types. This analysis includes a higher CAVEX than the source term previously analyzed in the current pH and AST LOCA analyses. The bounding dose results presented in the LAR are based on this bounding CAVEX source term. The licensee states in the LAR:

The revised analysis was performed in accordance with Regulatory Guide (RG) 1.183 (Reference 6.6) to confirm compliance with the acceptance criteria in 10 CFR 50.67.

The NRC staff observes that this approach would include the limitation in RG 1.183, Revision 0, Table 1 (footnote 10), which bounds the peak burn up of the fuel up to 62,000 MWD/MTU as stated in Section 4.3, Release Fractions and Timing, in LAR, Enclosure B.

Because the proposed inventories were derived from the ORIGEN-2 code, and are based upon:

(1) bounding fuel inventories, (2) bounding fuel enrichments, (3) the DNPS rated thermal power (with ECCS) uncertainty, (4) equilibrium core inventory values, and (5) the licensees statement in the LAR that the revised analysis was performed in accordance with RG 1.183, Revision 0, the NRC staff concludes that the core inventory used for the design basis LOCA radiological analysis is consistent with RG 1.183, Revision 0, section 3.1, Fission Product Inventory, and is acceptable for use in the DBA LOCA radiological analysis.

Further, the NRC staff found that the modeled AST/core inventories, BWR core inventory fractions and release timings, DNPS maximum rated thermal power with uncertainty, and the chemical forms and percentages of radioiodine released into the containment were based on bounding fuel inventories and enrichments, and are consistent with the guidance in RG 1.183, Revision 0, and, therefore, are acceptable.

3.1.2.2 MSIV Leakage In the LAR, the licensee states that the MSIVs are postulated to leak at a total design leakage rate of 250 scfh for Unit 2, and 350 scfh for Unit 3. Unit 3 is analyzed for a higher leakage rate than Unit 2 because the Unit 3 MSIV to CR ground-level release /Q values are lower than the

/Q values for Unit 2. The radiological consequences from postulated MSIV leakage crediting drywell sprays to reduce isotopes escaping containment, is analyzed and combined with the radiological consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the design basis LOCA as summarized in , Table 3-2, LOCA Dose Consequence Summary, of the LAR. Table 3-2 shows that the CR dose corresponds to a design leakage of 250 scfh from Unit 2, EAB and LPZ doses correspond to a design leakage of 350 scfh from Unit 3, and the CR doses from a release from Unit 3 are lower than the Unit 2 doses due to crediting the lower /Q value from a Unit 3 release from the MSIVs to the CR.

The LAR stated that all MSL piping sections between the RPV nozzle and outboard MSIVs used in the MSIV leakage release paths are assumed to remain intact and can perform their safety function during and following a safe shutdown earthquake (SSE). Based on the structural integrity and functional performance of the MSL piping up to the outboard MSIV to withstand the SSE, the horizontal pipe surface area and volume is credited in the aerosol removal calculation.

A total MSIV leakage of 250 scfh for Unit 2, and 350 scfh for Unit 3, was assumed to occur:

1) 100 scfh through the steam line with the failed MSIV. The failure is assumed to cause a single MSL to have a disproportionately high flow to artificially increase the total allowed MSIV leakage. The steam line with the failure is the shortest of the four steam lines, so increasing the flow rate in this steam line reduces the overall credited aerosol and elemental iodine removal. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes, are credited in the steam line between the RPV nozzle and outboard MSIV.

2) 100 scfh through first intact steam line. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes, are credited in the steam line between the RPV nozzle and outboard MSIV.

3) 50 scfh through the second intact steam line. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes are credited in the steam line between the RPV nozzle and outboard MSIV.

4) 0 scfh through the fourth steam line for both units.

The NRC staffs evaluation of these assumptions is provided below in Section 3.1.2.3.

3.1.2.3 LOCA Analysis LAR Attachment 1, Table 3-1: Summary of LOCA Analysis Revisions, provides a summary of the changes and the reason for each change to the methodology and inputs of the revised design basis LOCA radiological analysis compared to the CLB analysis.

LAR Attachment 1, Table 3-2: LOCA Dose Consequence Summary, shows that the CR dose remained relatively stable compared to the CLB CR dose even with the increase in MSIV leakage. This primarily results from crediting drywell sprays to reduce isotopes escaping containment. This issue is discussed below, in the subsection entitled Drywell Spray Assumptions in the LOCA Model.

The revised design basis LOCA radiological analysis credits the reduction in the containment leakage and MSIV leakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50 percent of the maximum leakage with respect to RADTRAD flowrates.

The radiological consequence analysis for the postulated design basis LOCA in the CLB is based on the NRC-approved use of an AST described in the DNPS UFSAR, Section 15.6.5, Loss-of-Coolant Accidents Resulting from Piping Breaks Inside Containment. The results of the offsite and CR doses were re-analyzed to demonstrate that the engineered safety features (ESFs) designed to mitigate the radiological consequences at DNPS, Units 2 and 3, will remain

acceptable after the implementation of the increased MSIV leakage rate limits and crediting the drywell spray system for accident mitigation to support the TS changes requested in the LAR.

In its evaluation of the design basis LOCA radiological analysis, the licensee included dose contributions from the following activity release pathways:

Containment leakage, ESF leakage, and MSIV leakage The licensee included the following design basis LOCA dose contributors to the CR analysis:

Reactor building shine, External cloud shine, and CR filter shine In the licensees letter dated May 6, 2020, Attachment 2, Table 8-4, Bounding LOCA doses using GNF3 fuel with 250 scfh leakage for Unit 2 and 350 scfh leakage for Unit 3, the licensee showed that the CR TEDE dose (5.00E+00) is the most limiting of the three receptor locations (CR, EAB, LPZ). MSIV leakage accounts for approximately 80 percent of the total dose to the CR. The next highest contributor of dose is External Cloud Shine, contributing approximately 11 percent of the CR dose. Therefore, the NRC staff focused its review of the MSIV leakage increase proposed in the LAR by reviewing the MSIV leakage release pathway, which is the most significant contributor to the design basis LOCA dose.

The licensee calculated and submitted the results of the offsite and CR doses and provided the major assumptions and parameters used in its dose calculations in the LAR. The licensee stated in the LAR that after implementing the increased MSIV leakage rate limits, crediting the drywell spray system for accident mitigation, making the TS changes, using an AST, and crediting the existing ESF systems at DNPS, that the radiological consequences of the postulated design basis LOCA at the EAB, LPZ, and in the CR will meet the dose criteria specified in 10 CFR 50.67 and in 10 CFR Part 50, Appendix A, GDC 19. The NRC staff evaluated the licensees assessment and agrees with these conclusions.

The revised design basis LOCA radiological analysis provided in the LAR removed credit in the CLB for the Powers natural deposition model from NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments (ADAMS Accession No. ML100130305), dated July 1996, in the drywell for the particulate (aerosol) deposition/plateout in containment. This consisted of approximately 32,250 square feet (ft2) of surface area available in the drywell for deposition and plateout of aerosols not credited in the RADTRAD model. The only surface area credited for deposition and plateout was upstream of the outboard MSIVs in three of the four MSLs which has a surface area of 237 ft2. In the LAR supplement (dated May 6, 2020), Attachment 2, Section 5.8, Changes Between Revision 4 and Revision 5, the licensee states that credit was removed for the Powers natural deposition model due to an error notice posted to the RADTRAD Industry Users Groups website (radtrad.com) indicating that it may underestimate doses for BWRs. The NRC staff found this acceptable because the Powers deposition model was not used in the revised design basis LOCA radiological analysis.

Drywell Spray Assumptions in the LOCA Model:

RG 1.183, Revision 0, Appendix A, section 3.3, Timing of Release Phases, states that reduction in airborne radioactivity in the containment by containment spray systems that have been designed and are maintained in accordance with Chapter 6.5.2, Revision 4, of the SRP, may be credited in dose consequence analyses. In the LAR, the licensees calculation (DRE05-0048) credits a spray removal rate based on a spray pump with a volumetric flow rate of 2,352 gpm.

The LAR analysis assumed that the spray would be initiated by manual action in 10 minutes post-accident with an assumed termination at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and a fall height of 27 5 (8.36 meters (m)) based on the difference in elevations between the lower drywell spray header and the bottom of the drywell floor.

The NRC staff reviewed the DNPS UFSAR, Section 6.2.2, Containment Heat Removal Systems, to determine if the containment spray systems were designed to provide a reduction in airborne activity consistent with SRP Section 6.5.2. Based on the NRC staffs review, it appeared that the spray systems were designed for pressure reduction and not specifically for reducing airborne radioactivity. The NRC staff noted that containment spray design requirements regarding the ability to reduce airborne radioactivity were discussed in LAR, Enclosure B, Section 2.1.3, Reduction in Airborne Activity Inside Containment, in a comparison between SRP Section 6.5.2 review items and how those items are addressed in the revised analysis. The NRC staff requested the licensee to provide additional information describing how the design characteristics of the containment spray systems provide a reduction in airborne radioactivity as discussed in Enclosure B, Section 2.1.3, of the LAR. The request for additional information also asked the licensee to describe how taking credit for reduction in airborne radioactivity with the use of sprays is consistent with SRP 6.5.2, and how this change will be incorporated into the DNPS UFSAR.

The NRC staff reviewed the licensees calculation of the particulate removal coefficient as documented in LAR, Enclosure B, Section 7.11, Spray Calculations, (pg. 64). Based on this review, it appeared that the spray drop-fall height of 8.36 m (27.4 ft) was determined by the difference in elevations between the lower drywell spray header and the bottom of the drywell floor. This method did not appear to consider the obstructions that are present in the drywell, which could reduce the effective spray drop fall height. In addition, the analysis assumed a spray flow rate of 2,352 gpm. As with spray drop fall height, obstructions in the drywell could reduce the effective spray flow rate available for reducing airborne radioactivity. The NRC staff noted that both the unobstructed free fall height and spray flow rate are important factors in determining the ability of the containment sprays to effectively reduce airborne radioactivity.

NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, Section H, dated June 1993, (ADAMS Accession No. ML063480542), discusses the issue of obstructions interfering with the effectiveness of sprays as follows:

H. Droplet-Structure Interactions Reactor containment buildings are not simple, open volumes. Immediately below spray headers there is often a substantial open space. But, eventually, falling drops begin to encounter equipment, structures and operating floor of the reactor. The drywells of Mark I containments are well-known for the congestion that can interfere in the free fall of water droplets.

The flooring in many reactor containments is grating or so-called expanded sheet metal. Below the flooring are large volumes which, in a severe reactor accident, would hold aerosol-contaminated gas. It is of interest to know, then, if spray droplets, after hitting structures and the open flooring, would continue to sweep aerosols from the containment atmosphere. Certainly, in the case of the design basis analysis of iodine removal from containment atmospheres, it has been traditional to assume droplets are ineffective once they have hit a structure or the flooring.

The NRC staff requested additional information from the licensee requesting a justification for the use of the assumed spray fall height of 8.36 m (27.4 ft) in the determination of the particulate removal coefficient, without considering obstructions present in the drywell. In the supplemental letter dated May 6, 2020, the licensee stated the spray removal coefficient used in Revision 6 of DRE05-0048 Dresden Unit 2 & 3 Post-LOCA EAB, LPZ and CR Dose - AST Analysis for a decontamination factor (DF) 50 is 15.0 hour-1 (hr1). The response stated that this spray removal coefficient value was based in part on a spray fall height calculated as the difference between the lower spray header elevation (529 ft. - 9 in. (inch)) and the bottom of drywell elevation (502 ft. - 4 in.). Consistent with SRP Section 6.5.2, this reduced value of 15.0 hr-1 for DF 50 was further reduced by a factor of 10 to 1.5 hr-1 when a DF of 50 was reached in the RADTRAD model. The NRC staff agrees with this reduction in spray removal coefficient, as discussed at the end of this section.

In its response, the licensee also noted that in addition to the equipment installed in the drywell, the obstructions included two floors of grating between the spray headers and the bottom of the drywell: one between the two spray headers at elevation 537 ft. - 11/4 in. and one below the lower spray header at elevation 515 ft. - 53/4 in. The licensee explained that since both elevations of spray nozzles (upper at 551 ft.-2 in. and lower at 529 ft.-9 in.) will be available following a LOCA, the average fall height between these two elevations could have been used to calculate the fall height, but using the lower header elevation provides some additional conservatism. The licensee concluded that in conjunction with the conservatisms in the spray flow rate discussed in its response, the overall spray removal coefficient was conservative as compared to the methodology used for Nine Mile Point, Unit 1 (NMP1) and Oyster Creek (OC),

which reduced the average spray header fall height to account for obstructions including grating and equipment.

In its letter dated May 6, 2020, the licensee also referred to the methodology used for the NMP1, OC, and Nine Mile Point, Unit 2 (NMP2), AST implementation, in which specific reductions in the spray removal coefficient calculation were made, based on obstructions in the drywell or blocked nozzles that may impede flow. For example, the NRC staffs SE associated with the implementation of the NMP1 AST states, in part, the following:

To account for drywell congestion, the licensee multiplied the secondary spray flow rate by 0.67 for additional conservatism. Also, the fall height of 21.4 feet, used by the licensee conservatively reflects a one-third reduction to account for drywell congestion.

The licensees response also referred to the OC AST, which applied the three-dimensional modeling of the drywell. This modeling resulted in a 33.3 percent reduction in fall height to account for obstructions and a 33.3 percent reduction in flow rate based on a modular accident analysis program analysis. The NRC staffs SE associated with the OC AST states, in part, that Drywell congestion explicitly addressed by reduced spray flow and fall height.

To address its concern related to spray flow rate, the NRC staff requested the licensee to provide a justification for the use of the assumed spray flow rate of 2,352 gpm in its determination of the particulate removal coefficient, which apparently did not consider obstructions present in the drywell. As noted above, in the licensees supplemental letter dated May 6, 2020, Attachment 2, DRE05-0048, Revision 6, Section 7.11, the licensee discussed using an assumed drywell spray flow rate of 2,352 gpm. This volumetric flow rate is based on 160 drywell spray nozzles providing 14.7 gpm each as described in the May 6, 2020, LAR supplement, Attachment 2, Section 5.2, Accident-Specific Design Inputs/Assumptions, (subpart 5.3.2.12 Containment Spray Parameters, pg. 30). Each ring header contains 160 nozzles spaced around the drywell. The licensees response further explained that the spray flow rate assumed in the analysis was based on only a single header providing flow even though both headers can be supplied simultaneously by a single LPCI pump. In addition, UFSAR, Section 6.2.1.3.3, states, in part, that the design basis drywell spray flow rate is 4,750 gpm and the wetwell spray flow rate is 250 gpm. Importantly, as noted in TS SR 3.6.2.3.2, each required LPCI pump develops a flow rate greater than or equal to 5,000 gpm while operating in the suppression pool cooling mode. This LPCI pump flow rate is substantially greater than the drywell spray flow rate of 2,352 gpm (~47 percent of the TS SR 3.6.2.3.2 LPCI pump flow rate) that was assumed in the licensees calculation of the spray removal coefficient.

In support of the LAR, the licensees response of May 6, 2020, referred to the NMP2 AST methodology (which the NRC staff approved for NMP2), that was used to determine the flow rate reduction due to potential nozzle blockage. This methodology assumed that certain percentages of nozzles were blocked using survey data. Based on the NMP2 AST methodology, the licensee determined that the difference in the overall reduction was negligible and will still lead to a spray removal coefficient greater than 15.0 hr-1.

The NRC staff notes that the overall spray removal coefficient considers both spray fall height and spray flow rate and there was no specific reduction taken in the spray fall height to consider obstructions in the drywell. However, based on a review of the licensees responses to the NRC staffs questions, as noted above, the NRC staff has concluded that the overall spray removal coefficient of 15.0 hr-1 for a DF 50 and 1.5 hr-1 for a DF > 50 is acceptable, because it addresses the overall impact of spray fall height and spray flow rate from obstructions that are present in the drywell which could reduce the effective spray drop fall height and effective spray flow rate available for reducing airborne radioactivity. The initial value of the spray removal rate, and the reduction in the spray removal rate by a factor of 10 is consistent with RG 1.183, Revision 0, RP 3.3. The RG and the SRP state in part the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The NRC staff concludes the licensees analysis remains consistent with applicable regulatory guidance, and, therefore, is acceptable.

Drywell Spray Credit for Deposition and Plateout in the LOCA Model:

(Crediting Iodine Removal in MSL Piping)

RG 1.183, Revision 0, Appendix A, Section 6.3, states, in part, that the Reduction of the amount of released radioactivity by deposition and plateout on steam system piping upstream of the outboard MSIVs may be credited, but the amount of reduction in concentration allowed will be evaluated on an individual case basis.

LAR, Attachment 1, Table 3-1, Summary of LOCA Analysis Revisions, presents changes to the CLB for the revised design basis LOCA radiological analysis. One of the proposed changes involves a change to the credit for elemental iodine removal in the MSLs. The CLB credits elemental iodine removal in the two intact steam lines but not in the line with the failed MSIV.

The LAR proposed to substantially increase the elemental iodine removal in the MSLs between the RPV and the outboard MSIV by crediting elemental removal in the line with the assumed failed MSIV and by increasing iodine removal in the previously credited MSL volumes from 50 percent to up to about 98 percent.

During the NRC staffs evaluation of LAR, Enclosure B, Section 7.3, Main Steam Line Volumes

& Surface Area for Plateout of Activity, (pg. 38), the staff identified certain discrepancies in the tabulated data and parameter values applied in the revised LOCA radiological analysis, which included:

In LAR, Enclosure B, Table 40, MSIV Failed & Intact Steam Line Volumes for Elemental Iodine Removal Efficiency Calculation, (pg. 97), the calculated volume for Note: D (Volume V4) given as 4.33 m3 should be 4.64 m3. The calculated volume of Note: E (Volume V5) of 4.33 m3 should be 1.39 m3. (The NRC staff noted the licensees actual table parameters were correct; the licensee characterized these discrepancies as typographical errors that were corrected in the licensees supplemental letter dated May 6, 2020).

In LAR, Enclosure B, Table 46, Elemental Iodine Deposition Rate - Intact Steam Line Volume V4, (pg. 100), the Main Steam Line Total Surface Area (B) given as 10.07 m2 (durations 0-480) should be 12.35 m2 (durations 0-480). This error impacted the Elemental Iodine Removal Rates (hr-1) and Elemental Iodine Deposition Efficiencies for all listed post-LOCA times in Table 46. (The licensee corrected the table parameters including a recalculation of the Elemental Iodine Deposition Efficiencies, in the licensees supplemental letter dated May 6, 2020).

In LAR, Enclosure B, Table 51, Net Elemental Iodine Removal Efficiency - Intact Steam Line Volume V4, (pg. 103), the Elemental Iodine Deposition Efficiencies, Elemental Iodine Resuspension Efficiencies, and Elemental Net Deposition Efficiencies (%) for all listed post-LOCA times were impacted by the above Table 46 discrepancies. (The licensee provided corrected parameters for Elemental Iodine Deposition Efficiencies, Elemental Iodine Resuspension Efficiencies, and Elemental Net Deposition Efficiencies

(%) in its supplemental letter dated May 6, 2020).

As a result of the Table 51 observed discrepancies, the RADTRAD model input parameter values for elemental iodine are impacted.

To address the discrepancies in the tabulated data and parameter values applied as inputs in the revised design basis LOCA radiological analysis, the staff requested additional information from the licensee and requested that further analyses be provided to address the impact on the calculated CR and offsite doses in the revised design basis LOCA radiological analysis. The licensees letter dated May 6, 2020, corrected the errors and made the following changes, some of which resulted in RADTRAD recalculations:

(Table 40) the typographic errors were corrected in the tables Note for the calculated volumes for D (Volume V4) and E (Volume V5). The correct volumes were provided

and were used to calculate the values in the RADTRAD models, and were found to have no impact on the calculation result, (Table 46) the Main Steam Line Total Surface Area (B) was corrected and used for durations (0-480), and the corresponding Elemental Iodine Removal Rate (D) and Elemental Iodine Deposition Efficiency (E) were updated. This change increased the removal of elemental iodine which resulted in a slight decrease in the calculated doses, and, (Table 51) the Elemental Iodine Deposition Efficiency (A) was updated for durations (0-480) to the corrected values from Table 46. The resulting Elemental Iodine Net Deposition Efficiency (C) was updated for each duration. The response stated that the changes did not result in any changes to the design basis dose consequences.

The NRC staff reviewed the licensees corrected information and finds it resolves the staffs concerns. Accordingly, the NRC staff concludes that the licensees response to correct the data and recalculate the impacted parameters to the correct values, is acceptable.

(Aerosol Removal in Steam Lines with Sprays Credited)

RG 1.183, Revision 0, Appendix A, Section 6.3, states, in part, that the Reduction of the amount of released radioactivity by deposition and plateout on steam system piping upstream of the outboard MSIVs may be credited, but the amount of reduction in concentration allowed will be evaluated on an individual case basis. Appendix A, Section 6.5 states, in part, that a reduction in MSIV releases that is due to holdup and deposition in main steam piping downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by off gas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). The amount of reduction allowed will be evaluated on an individual case basis.

LAR, Attachment 1, Evaluation of Proposed Changes, (pg. 17), states, in part:

The currently approved main steam line aerosol removal model (AEB-98-03) does not include deposition by thermophoresis, diffusiophoresis, or flow irregularities.

Therefore, it is reasonable to consider the use of aerosol removal by sprays and aerosol removal in the main steam lines as independent removal mechanisms because they rely on different physical mechanisms except for diffusiophoresis. However, neither the spray model nor the MSL aerosol removal model consider removal by diffusiophoresis making the model conservative with respect to the experimental data.

The NRC staff agrees with the licensee that the omission of these aerosol removal mechanisms in its dose consequence analyses is conservative.

LAR, Enclosure B, Section 5.8, Changes Between Revision 4 and Revision 5, (pg. 31), states, in part:

Drywell spray meets the requirements in NUREG-0800 Section 6.5.2 as demonstrated in Section 2.1.3 and has been previously accepted for Nine Mile Point Units 1 and 2, Oyster Creek, and Hatch.

In this regard, the NRC staff notes that when it approved the AST applications for NMP1, NMP2, OC, and Hatch, drywell spray credits were accepted on an individual case basis, which included considerations of the particular design and different conditions. Accordingly, the staffs approval of those applications does not necessarily support the acceptability of this LAR.

During its review of the DNPS LAR, the staff observed that the LAR proposes to credit sprays to remove fission products following a design basis LOCA, but it does not appear to adjust the MSL aerosol deposition from the impact of the sprays in the revised LOCA radiological analysis. LAR, Enclosure B, Table 2, Rate Constant (s) for Aerosol Settling in Main Steam Piping, (pg. 56) shows the same 40th percentile aerosol settling velocity (0.00081 m/s or 9.56 ft./hr.) in all control volumes, just as in the CLB, with no credit for sprays. When applying credit for sprays, the staff considers that the use of the CLB 40th percentile for aerosol settling velocity is non-conservative. This is due to the fact that larger particles and droplets are preferentially removed using sprays, thereby driving the particle distribution towards smaller particles, causing a lower settling velocity. This effect would cause more activity available for release. The sprays change the aerosol particle distribution curve on a time-dependent basis through each control volume that impacts its removal in the MSLs.

From the NRC staffs examination of the information submitted by the licensee, it appears that the revised design basis LOCA radiological analysis considers aerosol removal by sprays, and aerosol removal in the MSLs, as independent removal mechanisms. The staff notes, however, that regardless of the specific removal mechanisms involved, larger aerosol particles in the containment atmosphere will be preferentially removed, thereby making subsequent removal of aerosols by deposition in downstream piping more challenging.

To address the staffs concern regarding the amount of aerosol removal in the MSLs with sprays credited, the NRC staff requested the licensee to provide additional information and a justification as to why the proposed aerosol settling velocity and model to credit sprays in the DNPS design are consistent with RG 1.183, Revision 0. In the licensees response, dated May 6, 2020, the licensee asserted that there are inherent conservatisms in the design basis LOCA radiological consequence assessment, and it presented sensitivity analyses to show the uncertainty introduced by the drywell spray effects on the aerosol deposition model. More specifically, the licensee stated that the DNPS CLB includes several conservatisms in the LOCA dose consequence assessment that were credited as part of the NRCs approval of the DNPS AST amendment, whose design basis was provided by DRE05-0048, Revision 1 (Attachment 2 to letter of May 6, 2020). The licensee further stated that both the CLB and the revised LOCA AST dose analysis (LAR supplement dated May 6, 2020) assume the drywell is the source of MSIV leakage in accordance with the NRC guidance in RG 1.183, Revision 0, and that it is appropriate to consider radionuclide removal mechanisms in the drywell before release via the MSIV leakage pathway. The NRC staff agrees with this conclusion, which is consistent with Regulatory Position 6.1, Assumptions on Main Steam Isolation Valve Leakage in BWRs, which states: For the purpose of this analysis, the activity available for release via MSIV leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage.

3.1.2.4 Licensees Sensitivity Study - Use of 20-Group Method By letter dated May 6, 2020, in response to an NRC staff question, the licensee stated that a simplified model was developed using first principles as identified in NUREG/CR-5966. The ordinary differential equation shown in NUREG/CR-5966 (pg. 1) was solved to provide an analytical solution of the suspended aerosol mass in the drywell. The spray removal rate in the simplified model is the same as that identified in LAR, Attachment 2, DRE-05-0048, Revision 6, Section 2.1.3, and RG 1.183, Revision 0, Appendix A, Section 3.3. Since sprays will remove aerosols at different rates depending on their particle size, the spray removal rate was adjusted by collection efficiency variation as provided in NUREG/CR-5966 (Figure 19).

The staff notes that NUREG/CR-5966 provides details on how sprays impact aerosols. The document indicates that sprays shift the sizes of aerosols in containment towards those that are removed most slowly (i.e., the mean aerosol size decreases as the sprays operate). Estimates of aerosol deposition in the steam lines is determined using, in part, Equation 5 of AEB 98-03, which indicates that aerosol settling is proportional to the square of the diameter of the aerosols.

Because the sprays shift the size of the aerosols to smaller sizes, the sizes of aerosols settling in the steam lines would decrease due to the prevalence of smaller diameter aerosols.

In the licensees sensitivity study, the suspended aerosol mass was solved from the beginning of the accident through the termination of the sprays at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for 20 distinct particle size groups. The mass of particles in each group is defined by the probability distribution associated with the source distribution.

The licensees response stated that the analysis assumed the size distribution of the particles released from the fuel was log-normal with 2-micron Aerodynamic Mass Median Diameter (AMMD) (0.473-micron geometric mean diameter) with a Geometric Standard Deviation (GSD) of 2. The aerosol mass was calculated for each group independently with no consideration of particles interacting with one another, so agglomeration is not accounted for. The NRC staff agrees that not taking credit for agglomeration is conservative. LAR, Attachment 1, of the letter dated May 6, 2020, Table RAI-2a: Drywell Particle Size Distributions, summarized the results of the 20-group particle size distribution in the drywell. The same attachment, in Figure RAI-2a, Time-Dependent Aerosol Particle Size Distribution, shows the time-dependent nature of the aerosol particle size distribution and the effect of the drywell spray in reducing the size of the particles in the model. The particle size and settling velocity distributions were then used to recalculate the aerosol removal factors using the equation found in LAR, Attachment 2, Section 7.4.1, of DRE-05-0048 (also summarized in LAR, Attachment 1, Table RAI 2d, Steam Line and Condenser Aerosol Removal Factors). The aerosol removal factors including spray, combined with the nodalization adjustments described below, are represented by the Base Sensitivity Case in Attachment 1, Table RAI-2e, Sensitivity Study Results. The sensitivity analyses use the Unit 2 RADTRAD model inputs. The licensees response stated that the relative changes in the calculated doses are expected to be similar for the Unit 3 RADTRAD model inputs.

In its letter dated May 6, 2020, the licensee stated that a total of seven additional sensitivity cases (S1 through S7) were performed using various combinations of breathing rate, MSIV impaction, and condenser holdup/aerosol deposition by varying the Base Sensitivity Case (base case). The base case is the Unit 2 DRE-05-0048, Revision 6, analysis. The Unit 2, DRE-05-0048, nodalization was modified to separately model each of the four MSLs shown in LAR, Attachment 1, Figure RAI-2b, Modified Nodalization for a Single Steam Line. As a result, each sensitivity case included four RADTRAD models, one for each line with three well-mixed

nodes per line. The licensee asserted that the outboard steam line up to the turbine stop valve (TSV) at DNPS are seismically qualified, so including holdup and deposition in this piping as part of the outboard compartment (third well-mixed node in Figure RAI-2b) conforms with the guidance in RG 1.183, Revision 0. The data used to calculate the steam line and condenser aerosol removal rates are provided in LAR, Attachment 1, Tables RAI-2b and 2c, and are duplicated from DRE-05-0048, Sections 7.2 and 7.3. The sensitivity case results are summarized in LAR, Attachment 1, Table RAI-2e, Sensitivity Study Results.

The NRC staff noted that the basis for the 2-micron AMMD particle size and the methodology for the 20-group particle size distribution were not fully described in the licensees response to the staffs questions. Rather, LAR, Attachment 2, DRE-05-0048, Revision 6, Section 4.6.10, states, in part, For additional information on the models, data, and results refer to DRE20-0003, [AST LOCA Aerosol Removal Factors and Margin Assessment]. The licensee did not submit the calculations associated with the 2-micron AMMD particle size nor the 20-group method on the docket. Therefore, NRC staff did not review and evaluate these assumptions because: (1) no basis was provided for these assumptions, and (2) these assumptions were not used in the licensees proposed analysis of record.

3.1.2.5 NRC Staff Evaluation of the Licensees Sensitivity Analysis By letter dated May 6, 2020, in response to a NRC staff question, the licensee stated that a sensitivity analysis was performed to evaluate the impact of sprays on the aerosol settling velocity and to identify other inputs with well-defined uncertainty or conservatism that could be used to offset the uncertainty associated with the current aerosol deposition model. The staff evaluated each of these conservatisms and the associated sensitivity studies. As described below, the NRC did not consider in this evaluation six of the seven purported conservatisms. The staff agrees, however, that the last conservatism asserted by the licensee, associated with condenser hold up, is conservative, as discussed below and in section 3.1.2.7 of this SE, NRC Staff Risk and Engineering Insights. The licensee did not submit any of the sensitivity analysis as an analysis of record, therefore the NRC did not evaluate these studies as a basis for approval of the LAR.

As discussed below, a total of seven sensitivity cases were performed from various combinations of breathing rate, MSIV impaction, and condenser holdup/aerosol deposition by varying the base case. As shown in LAR, Attachment 1, Table RAI-2e, the base case indicates that the conservative modeling of the drywell spray on the aerosol removal in the MSLs, without making other adjustments, results in increased doses. The NRC staff notes that the base case results indicate that while the calculated CR dose exceeds the 5 rem TEDE acceptance criterion, the off-site doses to members of the public remain within the dose acceptance criteria.

The NRC staff also notes that the base case was produced for the purpose of conducting a sensitivity analysis and is not intended to replace the accident analysis of record. The accident analysis of record is the revised design basis LOCA radiological analysis as discussed in the letter dated May 6, 2020, in response to NRC staff questions (RAI No. ARCB-RAI-1, concerning obstructions in the drywell and RAI No. ARCB-RAI-2, concerning data and parameter discrepancies and the containment pathway added in the RADTRAD model). The analysis of record (AOR) did not utilize the sensitivity analysis, and as stated above, the NRC staff has not relied upon the sensitivity analysis in determining to approve the LAR. The NRC staff has concluded that the accident analysis of record demonstrates that the dose consequences of a design basis LOCA comply with all the applicable dose acceptance criteria and the regulatory requirements of 10 CFR 50.67(b)(2)(iii).

By letter dated May 6, 2020, the licensee asserted that the seven items listed below, could be used to address reduced aerosol removal rates due to drywell sprays, to provide conservatism in the licensees AST LOCA model. The NRC staffs evaluation is provided below under each of these asserted conservatisms. The NRC staff did not review or rely upon the sensitivity studies in its evaluation and, as discussed below, took into consideration only one of the licensees seven asserted conservatisms.

(1) Credit full drywell spray lambdas (not included in the LAR evaluation):

As stated in the LAR, credit for full drywell spray lambdas (representing removal rate per unit time) was not included in the licensees evaluation. Because an evaluation and technical basis were not included in the sensitivity study, the NRC staff did not review the assertion that credit for full drywell spray lambdas is a conservatism in the AST LOCA model.

Therefore, the NRC staff does not consider credit for full drywell spray lambdas to be a conservatism in the AST LOCA model.

(2) Credit for plateout and deposition in drywell (not included in the LAR evaluation):

As stated in the LAR, credit for plateout and deposition in the drywell was not included in the evaluation. Because an evaluation and technical basis were not included in the sensitivity study, the NRC staff did not review the assertion that the proposed models lack of credit for plateout and deposition in the drywell is a conservatism in the AST LOCA model. Plateout and deposition in the drywell would impact the aerosol distribution removal in the MSLs.

However, the LAR does not consider the impact of plateout or settling in the drywell on the credited setting in the MSL. Therefore, the NRC staff does not consider credit for plateout and deposition in the drywell to be a conservatism in the AST LOCA model.

(3) Inclusion of all four main steam lines for holdup and deposition:

Proposed SR 3.6.1.3.10 specifies the maximum leakage rate through each MSIV leakage path and the maximum combined leakage rate for all four MSLs. Although the maximum leakage could exist in four MSLs simultaneously, it could also exist in fewer than four MSLs simultaneously. Because the dilution and deposition differ between MSLs, the assumed inputs and modeling of the allowed leakage can impact the postulated dose. RG 1.183, Revision 0, Section 5.1.3, states, in part, that the numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. 10 CFR 50.67, states, in part, that [t]he fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of design analyses or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Therefore, NRC staff has concluded that it is expected the modeling of three MSLs rather than four MSLs would result in a more conservative postulated dose, consistent with RG 1.183, Revision 0, Appendix A (With regard to radiological consequences, a large-break LOCA is assumed as the design basis case for evaluating the performance of release mitigation systems and the containment and for evaluating the proposed siting of a facility. Therefore, the NRC staff does not consider modeling all four MSLs for holdup and deposition to be a conservatism in the AST LOCA model.)

(4) Outboard main steam line piping holdup and deposition:

LAR, Attachment 1, of the letter dated May 6, 2020, describing Outboard Main Steam Line Piping, and in Table RAI-2e, Sensitivity Study Results, considered credit for holdup and deposition in the piping from the outboard MSIV up to the TSV [turbine stop valve]. In its response, the licensee stated, in part, that The outboard steam lines up to the turbine stop valve at DNPS are seismically qualified, so including holdup and deposition in this piping as part of the outboard compartment (third well-mixed node in Figure RAI-2b) conforms with the requirements of RG 1.183. The licensees response did not state, however, that the TSV was seismically qualified, nor has the staff independently confirmed that the TSV is seismically qualified; therefore, the NRC staff does not consider credit for holdup in the section of piping from the outboard MSIV to the TSV to be a conservatism in the AST LOCA model. Accordingly, the NRC staff did not review and evaluate this as a conservatism in the AST LOCA model.

(5) More realistic CR operator breathing rate:

The LAR, Attachment 1 of the letter dated May 6, 2020, referenced breathing rate data from the Environmental Protection Agency (EPA)/600/R-09/052F, Exposure Factors Handbook:

2011 Edition, Table 6-17. Table 6-17 provides breathing rates as a function of age for various percentiles up to a maximum value. RG 1.183, Revision 0, provides a method acceptable to the NRC staff for demonstrating compliance with 10 CFR 50.67 and uses a constant value of 3.4E-04 m3/s (cubic meters per second) for the duration of the CR dose consequence analysis. By LAR supplement dated May 6, 2020, the licensees response to a staff question, included a recommended breathing rate for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by reduced breathing rates as referenced by the EPA handbook, of 3.28E-04 m3/s from 2 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 3.06E-04 m3/s from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 30 days. The licensee stated that the analysis considered the 95th percentile values from the EPA Handbook as light intensity work, which is typical of a CR operator. As a result, the observed CR dose was reduced when compared to the base sensitivity case.

The NRC staffs examination of the sensitivity cases (S1, S4, S6, and S7) when compared to the base sensitivity case (S0) as referenced in LAR, Attachment 1, Table RAI-2e, Sensitivity Study Results, shows that consideration of a more realistic CR breathing rate would result in a reduction (about 5 percent) of the CR dose (based upon the differences in dose results of Sensitivity Cases S0 and S1). The NRC staff notes, however, that while the use of a breathing rate for light intensity work might be justified during time periods of normal working conditions, it is not considered justified for determining design basis radiation exposures from access to and occupancy of the CR under accident conditions when CR operators would be expected to be at a higher level of stress and increased activities. Therefore, the NRC staff does not consider that the licensees use of a reduced breathing rate as a more realistic CR breathing rate constitutes an acceptable conservatism in the AST LOCA model.

(6) Aerosol impaction on the first closed MSIV:

LAR, Attachment 1, Table RAI-2e, considered credit for aerosol impaction on the first closed MSIV in the RADTRAD model. As a result, the observed CR, EAB, and LPZ doses in the sensitivity analysis were reduced when compared to the Base Sensitivity Case.

The NRC staffs examination of the sensitivity cases (S2, S5, S6, and S7) when compared to the base sensitivity case (S0) in Attachment 1, Table RAI-2e, showed that consideration of MSIV impaction results in a reduction (up to about 31 percent) of the CR, EAB, and LPZ doses.

The licensee referenced the NMP1 AST LOCA licensing basis described in calculation H21C092, U1 LOCA w/LOOP, AST Methodology, [Non-proprietary], (ADAMS Accession No. ML070110240), that credits the phenomenon of impaction at the first closed MSIV. The objective of this NMP1 calculation is to determine the offsite and CR radiological consequences arising from a LOCA at NMP1. The NMP1 analysis was conducted in accordance with 10 CFR 50.67 and RG 1. 183, Revision 0. The licensee stated that in this scenario, some of the aerosol particles would be deposited on the MSIV sealing surface as the aerosols entrained with the carrier gas leak through the closed MSIV. NMP1 conservatively determined that this impaction results in a DF of 2, which is modeled as a 50 percent filter in the transfer pathway through the first closed MSIV. This reduction is only accounted for once in each MSL. The licensee stated that this approach was approved for NMP1 and is reasonable given that the aerosol settling rates calculated in this sensitivity analysis are conservative and lower than those used in the cited analysis.

The NRC staff acknowledges that NMP1 included the following assumption in its MSIV leakage dose consequence analysis:

From calculation H21C092, as referenced above, NMP1 stated in Assumption 7, that aerosol reaching the first closed valve in [Reactor Building] RB bypass pathways (including MSIV leakage) experiences a DF of 2 due to impaction.

The NRC staffs SE associated with NMP1 dated December 19, 2007, stated:

The NRC staff believes that, though there is merit to this plugging phenomenon and impaction in theory, there is not enough empirical evidence, directly related to the unique and hypothetical conditions associated with a design-basis LOCA event, to warrant full credit for such a considerable DF attributable to impaction.

Therefore, the NRC staff does not generally endorse taking credit for impaction when modeling removal of particulates in main steam lines following a LOCA.

However, the NRC staff does believe that enough evidence exists to verify the conservatism of a DF of 2 in the specific design-basis LOCA model at [Nine Mile Point Unit 1] NMP1. The contribution of this impaction DF to the overall iodine activity decontamination, does not lead to an excessive overall credit for iodine removal in the MSLs. Based on the approximate DF of 4 that credits for removal by sedimentation (see Section 3.2.1.2.1.4), combined with this DF of 2, the licensee is assuming less than a 90 percent overall iodine removal efficiency in the steam lines. If this MSIV leakage pathway were modeled using a well-mixed model, as described and previously approved in AEB 98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1465) Source Term, December 9, 1998, the calculated activity removal in the MSLs would be analogous to that calculated by the licensee.

Therefore, the NRC staff finds the overall iodine removal credited by the licensee to be acceptable, as modeled for NMP1.

The NRC staff notes that the above excerpt from the NRC staffs NMP1 SE clearly states that it does not generally endorse taking credit for impaction when modeling removal of

particulates in MSLs following a LOCA. The NRC staffs SE concluded, however, that the credit for impaction and the overall iodine removal credit modeled for NMP1 were acceptable. This conclusion was based upon specific characteristics of the NMP1 analysis and should not be interpreted as an NRC staff acceptance of credit for impaction when modeling the removal of particulates in MSLs following a LOCA. Therefore, the NRC staff does not consider credit for MSIV impaction to be a conservatism in the DNPS AST LOCA model.

(7) Condenser holdup and deposition.

The licensee states in the LAR, a further conservatism that is not currently modeled in DRE-05-0048 is the holdup and aerosol deposition provided by the condenser. The LAR posits, depending on the event scenario, that multiple pathways could exist to route activity to the condenser, including the drain lines and the turbine itself.

The licensee states in the LAR, that the sensitivity analysis modeled an MSIV leakage pathway to the condenser through the drain from the MSL piping between the MSIVs. The licensee further states that this model neglects any holdup and deposition in the outboard MSL piping and that modeling the release to the condenser from the piping between the MSIV is consistent with other plants in the Exelon fleet (e.g., LaSalle and Limerick). The licensee states that operating experience associated with the North Anna earthquake and post-Fukushima evaluations have shown that components and piping systems typically used in this release path are sufficiently rugged to ensure they are capable of performing some level of radioactivity removal during and following an SSE. The licensee concluded it is reasonable to assume that the condenser pathway could be made available for mitigating the consequences of MSIV leakage. In the LAR, Attachment 1, Tables RAI-2b and 2c, and Attachment 2, DRE-05-0048, Sections 7.2 and 7.3, the licensee specifies the data used to calculate the steam line and condenser aerosol removal rates.

The NRC staffs examination of the sensitivity cases (S3, S4, S5, and S6) compared to the base sensitivity case (S0) in Attachment 1, Table RAI-2e, showed that consideration of condenser credit results in a reduction (up to about 95 percent) of the CR, EAB, and LPZ, doses.

The licensees sensitivity results demonstrate that the condenser is very effective in substantially reducing the dose consequences from MSIV leakage. While other elements assessed in the licensees sensitivity analysis provided relatively small decreases in the calculated doses, the evaluation of the condensers mitigation properties provided a substantial dose reduction. The NRC staff agrees that holdup and deposition in a main condenser, if available, would substantially reduce dose consequences during a postulated accident. (see Section 3.1.2.7).

In addition, RG 1.183, Revision 0, Appendix A, describes assumptions for evaluating the radiological consequences of a LOCA. Section 6 of Appendix A describes assumptions for MSIVs in BWRs. Specifically, assumption 6.5 states:

A reduction in MSIV releases that is due to holdup and deposition in main steam piping downstream of the MSIVs and in the condenser, including the treatment of air ejector effluent by off gas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a SSE. The amount of reduction allowed will be

evaluated on an individual case basis. References A-9 [J.E. Cline, MSIV Leakage Iodine Transport Analysis, Letter Report dated March 26, 1991, (ADAMS Accession No. ML003683718)] and A-10 [USNRC, Safety Evaluation of GE Topical Report, NEDC-31858P (Proprietary GE report), Revision 2, BWROG [Boiling Water Reactor Owners Group] Report for Increasing MSIV Leakage Limits and Elimination of Leakage Control Systems, September 1993, letter dated March 3, 1999, [ADAMS Accession Number 9903110303.] provide guidance on acceptable models.

The dose acceptance for the LOCA analysis described in RG 1.183, Revision 0, is based on the regulatory acceptance criteria for what is commonly termed the maximum hypothetical accident (MHA) or the maximum credible accident. The MHA is an accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible.

The fission product release assumed for an MHA evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite dose consequences evaluated based on design-basis containment leakage.2 Finally, the NRC staff notes that, by design, a nuclear power plant must be able to withstand an SSE, and an SSE would, therefore, not result in core damage. However, the design basis radiological assessment of an MHA deterministically imposes a fuel melt source term into the containment to test the ability of the plant to meet dose acceptance criteria. Since the SSE would not cause fuel damage, the inclusion of a fuel melt source term in the dose analysis implies that two independent highly unlikely events could occur during the analysis period: an event resulting in substantial fuel melt followed by a significant unrelated seismic event. The NRC staff has concluded that the probability of this sequence of events is very low. This is further discussed below in Section 3.1.2.7.

3.1.2.6 Transport of Radioactivity in the Drywell In Section 2.1.2, of the LAR, the licensee addressed the transport of radioactivity in the DNPS drywell. The NRC staff considered the following guidance in evaluating the licensees consideration of this matter:

2 The licensee asserted a further conservatism that was evaluated by the staff in the DNPS AST LOCA model.

Specifically, the licensee asserted that CR atmospheric dispersion (/Q) factors (values) have readily defined uncertainty distributions which, if incorporated in the analysis, would demonstrate there is a substantial amount of margin in the input parameters. The licensee further stated that for simplicity, the distribution of potential values for such input parameters were not included in the sensitivity study. The NRC staff notes that the use of atmospheric dispersion factors known as /Q values (effluent concentration () divided by the source strength (Q) at a given distance and direction from the source) in design basis dose consequence analyses is a well-established practice.

Atmospheric dispersion values are based on the evaluation of site-specific meteorological data. These data are processed to provide values at the 95 percent confidence level ensuring there is reasonable assurance that the acceptance criteria will not be exceeded. Therefore, the NRC staff does not endorse the concept of including less conservative /Q values, as proposed by licensee here, as an element to be considered in dose consequence sensitivity analyses.

RG 1.183, Revision 0, Appendix A, Section 3.1, states, in part:

The radioactivity released from the fuel should be assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs [pressurized-water reactors] or the drywell in BWRs as it is released. This distribution should be adjusted if there are internal compartments that have limited ventilation exchange. The suppression pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell.

RG 1.183, Revision 0, Appendix A, Section 3.3, states, in part:

Evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray drops.

RG 1.183, Revision 0, Appendix A, Section 6.1, states, in part:

Activity available for release via MSIV leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage.

The staff observed that in LAR, Enclosure B, Section 2.1.2, Transport in Primary Containment, (pg. 8), the licensee states, in part:

For calculating the MSIV leakage flow rates between the drywell and the environment, the flow rate analysis is based on the total drywell volume during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the LOCA, and then the combined drywell plus suppression chamber air volume after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, at which time the containment volume is expected to become well mixed following the restoration of core cooling.

Further, LAR, Attachment 2, DRE05-0048, Revision 6, Section 7.2.3, MSIV Leakage During 2-24 hrs, (pg. 39) states, in part:

Two hours after a LOCA, the drywell and suppression chamber volumes are expected to reach an equilibrium condition and the post-LOCA activity is expected to be homogeneously distributed between these volumes. The homogeneous mixing in the primary containment will decrease the activity concentration and therefore decrease the activity release rate through the MSIVs. To model the effect of this mixing, the MSIV flow rate used in the RADTRAD model is decreased by calculating a new leak rate based on the combined volumes of the drywell and suppression chamber.

In the LAR, Enclosure B, DRE05-0048, Revision 5, Section 2.1.2, Transport in Primary Containment, (pg. 9) references Table 2 of AEB 98-03, which shows the dependence of radiological consequences on containment mixing conditions for the Perry Nuclear Power Plant (PPNP). However, the PNPP has a Mark III containment, which is significantly different than the Mark I containment at DNPS. These differences are not addressed in the LAR. Accordingly, the NRC staff did not consider this aspect of the LAR in its evaluation.

The licensee proposes in the LAR a significant change to the CLB transport modeling in primary containment by adding a compartment in the drywell to credit sprays and by crediting transport between the sprayed and unsprayed portions of the drywell. As a result, it was not clear to the

staff that the assumption of equilibrium conditions at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> applies between drywell and wetwell volumes. The proposed credit for sprays and the addition of the sprayed compartment decreases the activity in the drywell from the activity assumed in the CLB and, therefore, will create a difference in the modeled activity in the sprayed drywell compartment as compared to the activity in the wetwell.

From the NRC staffs review of LAR, Enclosure B, Attachment 13.1 - RADTRAD Output File DRE3CL395.o0 (GNF3 Fuel) starting on page 869 of the LAR, it appears that the I-131 activity concentrations for the sprayed and unsprayed portions of the drywell do not reach equilibrium conditions until more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> beyond the time when drywell sprays are assumed to terminate at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> post-accident for aerosol removal. Therefore, to address the transport of radioactivity in the drywell, the NRC staff requested additional information from the licensee to explain why the high flow rates necessary to create equilibrium conditions between the drywell and wetwell would exist for the time period past 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the DNPS design.

By letter dated May 6, 2020, the licensee explained that the assumption of equilibrium conditions between the drywell and wetwell is based on the steaming/condensing phenomenon associated with the restoration of core cooling at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The response explained that although the wetwell is not modeled separately in the containment leakage and MSIV leakage RADTRAD cases, the wetwell volume is used in the MSL flow rate calculations starting at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and crediting drywell sprays for airborne fission product removal does not change the well-mixed assumption.

RG 1.183, Revision 0, Appendix A, Section 3.3, states, in part:

The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.

The licensee stated that its RADTRAD modeling is based on separating the unsprayed and sprayed drywell volumes because the drywell sprays are assumed to cover less than 90 percent of the drywell volume. This is consistent with RG 1.183, Revision 0, RP 3.3, recited above.

Thus, since the drywell sprays are assumed to cover less than 90% of the volume, the licensee did not consider the containment to be a well mixed volume. Furthermore, the licensee explained that the RADTRAD modeling is intended to conservatively concentrate the airborne activity in the sprayed volume directly connected to the MSIV leakage pathways and to maximize dose, not to accurately reflect the thermal-hydraulic conditions that would be present in the drywell. Also, the discrepancy noted in the (iodine) I-131 inventory between the sprayed and unsprayed volumes is unrelated to the well mixed assumption between the drywell and wetwell at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is a byproduct of the conservative modeling inside the drywell.

Based on the NRC staffs evaluation of the licensees response, as discussed above, the staff concludes the licensee adequately applied the guidance in RG 1.183, Revision 0, RP 3.3, by not considering the containment building atmosphere to be a well mixed volume until after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> upon restoration of core cooling. Accordingly, the staff concludes that the licensees consideration of the transport of radioactivity in the drywell is acceptable.

3.1.2.7 NRC Staff Risk and Engineering Insights The NRC staffs conclusions in this SE are primarily based on traditional deterministic review approaches since the licensees LAR was not submitted as a formal risk-informed submittal with PRA information in accordance with the guidance of RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256), dated January 2018.

The consideration of risk-informed principles is consistent with the Commissions direction in the SRM to SECY-19-0036, with the NRCs goal of becoming a more modern and risk-informed regulator and with the NRCs Principles of Good Regulation. Since this LAR is not a fully risk-informed submittal, the staff does not apply risk as the basis for acceptance of the request; however, the following risk and engineering insights inform the technical review by supporting the deterministic safety conclusions and enhance the NRC technical reviewers confidence in their safety determination.

The LAR states that aerosol holdup and deposition provided by the condenser is not modeled in DRE-05-0048 and that depending on the event scenario, multiple pathways could exist to route activity to the condenser, including the drain lines and the turbine itself. The licensee concluded that it is reasonable to assume that the condenser pathway could be made available for mitigating the consequences of MSIV leakage.

The NRC staff performed an independent assessment evaluating the capacity of the power conversion system (PCS) and condenser to accommodate a fission product holdup and retention volume for MSIV leakage resulting in dilution from radiological decay (half-life).

The NRC staff performed an independent probabilistic assessment evaluating the capability of the power conversion system (PCS) and condenser to accommodate a fission product holdup and retention volume for MSIV leakage. The NRC staff evaluated the seismic capacity of the SSCs in the PCS, including the main steam piping, equalization header, and condenser, to assess whether they would be available to accommodate a holdup volume for fission products following an SSE. The NRC staff used engineering considerations, such as plant operations and design information, as well as probabilistic and risk information, to complete the evaluation.

The NRC staff also considered more recent relevant operating experience, such as that obtained from the Fukushima Dai-ichi accident and the magnitude 5.8 earthquake that impacted the North Anna Power Station on August 23, 2011. The staffs independent assessment found it is reasonable to conclude that the SSCs in the PCS would be available following an SSE and that the likelihood that they would be unavailable to accommodate holdup volume for MSIV leakage is very low.

The staffs assessment provides an important insight in considering uncertainties in the calculation of the dose consequences of MSIV leakage. Specifically, the NRC staff recognizes there is a high probability that actual doses will be significantly lower than those estimated using deterministic methods that do not credit holdup and retention of the MSIV leakage within the PCS.

Based on the available information and its own independent assessment, using conservative assumptions about the seismic capacity of the SSCs in the release path, the NRC staff determined with a high level of confidence that the PCS will be available for fission product holdup, and retention, especially in the event of a design basis SSE. In addition, as mentioned in the Statement of Consideration for 10 CFR 50.67, defense-in-depth is addressed by using a

DBA in the deterministic dose calculation. Deterministic accident analyses are intentionally conservative in order to address uncertainties in accident progression, fission product transport, and atmospheric dispersion. Although probabilistic risk assessments (PRAs) can provide useful insights into system performance and suggest changes in how the desired defense in depth is achieved, defense in depth continues to be an effective way to account for uncertainties in equipment and human performance. The NRCs policy statement on the use of PRA methods (60 FR 42622; August 16, 1995) calls for the use of PRA technology in all regulatory matters in a manner that complements the NRCs deterministic approach and supports the traditional defense-in-depth philosophy.

Therefore, consistent with 10 CFR 50.67, the principles of risk-informed decision-making, and the Commissions direction to the staff (as cited), the NRC staff concludes that risk and engineering insights support the NRC staffs finding, based on its deterministic review, that there is reasonable assurance that the activities authorized by this amendment can be conducted without endangering the health and safety of the public.

3.1.2.8 Dose Consequence Conclusion As described above, the NRC staff reviewed the dose consequences of the maximum hypothetical accident to assess the radiological impacts of the proposed LAR at DNPS. The NRC staff concludes that the licensees analysis methods and assumptions are consistent with the regulatory requirements and guidance specified above in Section 2.3 of this SE. The NRC staff concludes, with reasonable assurance, based on the deterministic evaluation described in this SE, and the risk insights provided in Section 3.1.2.7 above, that the EAB, LPZ, and CR, doses comply with the cited regulatory acceptance criteria. Further, the NRC staff has determined that the LAR demonstrates sufficient safety margins with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters. Therefore, the NRC staff has concluded that the licensees proposed TS changes as discussed in this SE are acceptable with respect to the radiological consequences of a design basis LOCA.

3.1.3 Evaluation of Secondary Containment Drawdown Analysis LAR Enclosure A, DRE 19-0015, Revision 0a, Dresden Units 2 & 3 Secondary Containment Drawdown Analysis, documents the analysis of the DNPS RB pressure response following a design basis loss of coolant accident (LOCA) with a coincident loss of offsite power (LOOP) at DNPS. The licensee utilized RG 1.183, Revision 0, to support its analysis. Also, the licensee used the computer code GOTHIC version 8.2 to calculate the RB temperature and pressure response.

In the drawdown analysis, the normal secondary containment pressure, 0.25 w.g. vacuum, would rise above atmospheric pressure due to the isolation of the RB ventilation system. Then a single train of standby gas treatment system (SGTS) would draw the pressure down to the required minimum post-LOCA vacuum of 0.25 w.g. in 22.2 minutes.

The NRC staff performed confirmatory calculations, as follows:

a. RB Leakage Area In the staffs confirmatory calculations, the predicted airflow rate was converted to an equivalent or effective air leakage area, consistent with equation 33 in Chapter 26 of Reference 1,3 as follows:

AL = C5 Qr ( 2pr)1/2/CD Where:

AL = equivalent or effective air leakage area, in2 Qr = predicted airflow rate at pr, cfm = 2437 cfm (100% RB air change per day)

= air density, lbm/ft3 = 0.0705 lbm/ft3 @14.7 lb/in2, 100°Fahrenheit, and 20% Relative Humidity pr = reference pressure difference = 0.25 w.g.

CD = discharge coefficient = 0.6 C5 = unit conversion factor = 0.186 All the openings in the secondary containment boundary were combined into an overall opening area and discharge coefficient for the RB when the equivalent or effective air leakage area is calculated. One of the options is to set CD = 0.6 (i.e., the discharge coefficient for a sharp-edged orifice). The air leakage area of the secondary containment is, therefore, the area of an orifice (with an assumed value of CD) that would produce the same amount of leakage as the building envelope at the reference pressure.

AL = (0.186)(2437)[0.0705 (2x0.25)]1/2 0.6 = 283.7 in2 = 1.97 ft2 The NRC staffs drawdown calculation used a calculation result AL = 2.055 ft2.

b. Hydraulic Diameter In the staffs drawdown calculation, a hydraulic diameter of 36.6 ft, was used in GOTHIC modeling. This was used for the calculation of flow paths connecting the spent fuel pool flow boundary conditions to the refueling floor control volume.

The LAR, Enclosure A (Calculation No. DRE19-0015, Rev. 0), includes drawdown results using NRC-approved GOTHIC (v8.2) methodology. In its confirmatory analysis, the NRC staff evaluated the licensees use of the calculation for DH = 4A/PW, to calculate the hydraulic diameter, DH. The SFP area, A, is 1353 ft2. The wetted perimeter, Pw, is (33x2) + (41x2) =

148 ft. The hydraulic diameter, DH = 1353x4/148 = 36.6 (ft). The NRC staff concluded that the DH value, 36.6 ft, matches the value used in the DNPS calculation.

The NRC staff also performed confirmatory calculations on the external environment SGT resistance and the SFP evaporation rate. The staff concluded that the calculations used in the DNPS analysis match the staffs confirmatory calculations, and, are therefore, acceptable.

3 ORNL-NSIC-5, U.S. Reactor Containment Technology, A Compilation of Current Practice in Analysis, Design, Construction, Test, and Operation, Volume 2, 1965.

3.1.3.2 MSIV Leakage Rate The proposed leakage rate for SR 3.6.1.3.10 through each MSIV leakage path is 62.4 standard cubic feet per hour (scfh) for Unit 2 and 78 scfh for Unit 3 when tested at 25 pounds per square inch gauge (psig), and the combined leakage rate for all MSIV leakage paths is 156 scfh for Unit 2 and 218 scfh for Unit 3 when tested at 25 psig. The staff evaluated the rationale for the difference between the proposed leakage rate and the licensing basis leakage rate as follows:

a. Conversion Factor TS SR 3.6.1.3.10 converts the leakage at the design pressure of 43.9 psig to a TS leakage rate at the test pressure of 25 psig with the use of a conversion factor. According to the LAR, TS leakage rates are calculated using a conversion factor of 1.603.

The NRC staff used the laminar seepage adjustment factor formula to calculate the conversion factor value as follows:

Le / Lt = (Pe - 1/Pe)/(Pt - 1/Pt)

Where:

Le = leakage rate of containment system at design conditions, 1/time Lt = leakage rate of containment system at test conditions, 1/time Pe = containment absolute pressure, atmospheres (atm)

Pt = containment absolute pressure (test condition), atmospheres For DNPS:

Pe = (14.7 + 43.9)/14.7 = 3.987 atm Pt = (14.7 + 25)/14.7 = 2.701 atm Therefore, Le / Lt = [3.987 - (1/3.987)] / [2.701 - (1/2.701)] = 1.603.

The NRC staffs calculated conversion factor, 1.603, agrees with the value used in the DNPS LAR, and, therefore, that value is acceptable.

b. TS Bases Using the conversion factor of, 1.603, the leakage rate through each MSIV leakage path is 100 (= 62.4 x 1.603) scfh for Unit 2 and 125 (= 78 x 1.603) scfh for Unit 3, when tested at 43.9 psig.

The MSIV leakage rate assumed in the LOCA dose consequence analysis is 250 scfh and 350 scfh for Unit 2 and 3, respectively, at the peak calculated primary containment internal pressure for the design basis LOCA, Pa (internal pressure), of 43.9 psig.

In the LAR, Attachment 1, Table 3-1 Summary of LOCA Analysis Revisions, indicates that the total MSIV leakage of 250 scfh for Unit 2 and 350 scfh for Unit 3 is assumed to occur as follows:

Leakage of 100 scfh for Unit 2 and 125 scfh for Unit 3, occurs through the steam line with the failed MSIV; Leakage of 100 scfh occurs through the first intact steam line for Unit 2, and 125 scfh for Unit 3; Leakage of 50 scfh occurs through the second intact steam line for Unit 2, and 100 scfh for Unit 3; and, Leakage of 0 scfh occurs through the fourth steam line for both Units.

For Unit 2, the combined leakage rate is 250 scfm. For Unit 3, the combined leakage rate is 350 scfm. The calculated combined leakage rates support the Technical Specification basis for the change to SR 3.6.1.3.10. The NRC staff considers that the flow distributions were chosen to be consistent with the CLB analysis, and, therefore, are acceptable.

As described above, the conversion factor for MSIV leakage rate is calculated using formulas from a widely accepted methodology, ORNL-NSIC-5, U.S. Reactor Containment Technology, Oak Ridge National Laboratory and Bechtel Corporation, A Compilation of Current Practice in Analysis, Design, Construction, Test, and Operation, Volume II, dated August 1965, as referenced in the LAR, Section 2.1. Also, the NRC staff has determined that the MSIV leakage model, credited in the licensees revised analysis, is consistent with the CLB analysis, and, therefore, is acceptable.

Based on the NRC staffs review and evaluation of the licensees proposed MSIV leakage rate, including supporting calculations that were verified and/or confirmed by the staff, the staff concludes the changes to SR 3.6.1.3.10, as proposed, are acceptable.

3.1.3.3 SR 3.6.4.1.1, Secondary Containment The licensees AST analysis assumed that the required secondary containment vacuum would be reached in 25 minutes. Required Action A.1 of LCO 3.6.4.1, Secondary Containment, states, Restore secondary containment to OPERABLE status, with a Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed note to SR 6.6.4.1.1 would allow the LCO not to be met for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if an analysis demonstrated that one SGT subsystem could establish the required secondary containment vacuum. During normal operation, conditions may occur that result in SR 3.6.4.1.1 not being met for short durations. For example, occasional wind gusts may lower external pressure on secondary containment walls, or normal ventilation system component failures or realignments may affect the secondary containment vacuum. These conditions most often do not represent degradations of the secondary containment boundary or the SGT systems ability to accomplish its specified safety function.

The proposed Note to SR 6.6.4.1.1 permits confirmation of secondary containment operability by confirming that one SGT subsystem could accomplish the required drawdown. This provides a limited exception to meeting the SR acceptance criterion. While the duration of these occurrences are anticipated to be brief and infrequent, the allowance is permitted for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which is consistent with the time permitted for secondary containment to

be inoperable per Condition A of LCO 3.6.4.1. The NRC staff concludes that since LCO 3.6.4.1 will be met, the addition of the proposed Note to SR 3.6.4.1.1 is acceptable.

3.1.4 Environmental Qualification (EQ)

Based on the licensees proposed TS changes, as discussed Section 2.2.2 above, the NRC staff evaluated whether electrical equipment and components would remain bounded by the existing EQ due to the proposed changes.

In the LAR, the licensee stated that the EQ doses are not impacted due to the proposed change because the current EQ design basis does not include source term in the main steam lines downstream of the MSIVs. Additionally, the licensee is crediting the drywell sprays to mitigate the consequences of a DBA. However, the licensee did not provide an evaluation of the impact of the MSIV increased leakage rate and the effect of using the drywell sprays on the temperature, pressure, or humidity, electrical equipment. The NRC staff requested the licensee to provide additional information verifying that the temperatures, pressures, and humidity remain bounded by the existing EQ for electrical equipment and components impacted by the MSIV increased leakage rate and drywell spray. By letter dated May 6, 2020, the licensee explained that the bounding accident temperature and pressure profiles in the main steam tunnel and turbine building are associated with a high energy line break (HELB) in the main steam tunnel.

When the increased MSIV leakage is considered, the HELB temperature and pressure profile in these zones continues to bound the LOCA profile. Additionally, the accident humidity in these zones is already assumed to be 100 percent. Therefore, the proposed increase in allowable MSIV leakage and credit for drywell spray would not contribute additional environmental impact to equipment qualified for use in the main steam tunnel or the turbine building.

Based on the staffs review of the licensees response, the NRC staff concludes that the licensee adequately demonstrated that the temperature, pressure, and humidity resulting from the proposed changes remain bounded by the existing EQ program for electrical equipment, as a result of the proposed changes and, therefore, the EQ of electrical equipment under the LAR is acceptable.

Additionally, during the NRC staffs review of the LAR, it was unclear to the staff as to whether the licensee considered the impact of the proposed change on nonsafety-related equipment whose failure under postulated environmental conditions could prevent the satisfactory accomplishment of safety functions by the safety-related equipment. Therefore, the staff requested additional information to explain how they assessed the impact of the proposed change on nonsafety-related equipment whose failure under postulated environmental conditions could prevent the satisfactory accomplishment of safety functions by the safety-related equipment. The licensee stated in its response (May 6, 2020) that since there is no change to EQ design basis temperatures, pressure, humidity, or radiation values, the proposed increase in MSIV leakage has no impact on nonsafety-related equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions by the safety-related equipment. Since the proposed LAR will not result in a change to existing EQ design parameters (i.e., temperature, pressure, radiation, humidity, etc.),

the NRC staff concludes that the proposed TS changes will not cause nonsafety-related electrical equipment to prevent satisfactory accomplishment of safety functions, and, therefore, the TS changes are acceptable.

The NRC staff also requested that the licensee confirm whether any components are being added to the EQ equipment list to comply with 10 CFR 50.49 due to the proposed change. In its response (May 6, 2020), the licensee stated that no components are being added to the EQ equipment list due to the proposed increase in allowable MSIV leakage. Following an evaluation of the licensees written responses, the NRC staff concludes that the licensee has adequately established that that no new electrical equipment is required to be added to the licensees 10 CFR 50.49 EQ program as a result of the proposed change to the MSIV leakage rate.

Based on the NRC staffs evaluation of the LAR and the licensees responses to staff requests for additional information, as it pertains to the proposed TS changes, and since the DNPS EQ program has been found to be consistent with the regulatory requirements of 10 CFR 50.49, the NRC staff have concluded the licensees TS changes, as proposed, are acceptable.

3.1.5 Credited Operator Actions (Human Factors)

Based on the licensees proposed TS changes, as discussed in Section 2.2 above, the LAR credits operator actions associated with the manual initiation of drywell spray during some accident conditions. There are no changes to the design of the human-system interfaces (alarms, controls, and displays) as a result of this LAR.

Description of the Credited Operator Action In Section 2.2 of the LAR, New TS 3.6.2.6 Drywell Spray, the licensee stated that operator actions associated with the use of drywell sprays are credited to reduce airborne activity in the drywell. The drywell sprays are expected to mitigate the consequences of a postulated LOCA by scrubbing radionuclides from the air in the drywell. The licensee also stated that the CLB does not credit the associated operator actions. By letter dated August 24, 2020, in the licensees response to a NRC staff question, the licensee stated that operator actions to initiate the drywell spray were previously included in the Operator Response Time Program at Dresden for some accident conditions, some of which have time restrictions that are more conservative than the accident conditions assessed for this LAR. The licensees response also described the procedures and training used to support CR operators. In addition, the licensee stated that the Operator Response Time Program at Dresden procedure will be updated to ensure that the drywell initiation actions are appropriately included.

The newly credited operator actions are very similar to operator actions which have been tracked in the Operator Response Time Program at Dresden procedure. The main difference between the newly credited actions and the existing operator actions is the purpose for performing these actions (removing heat as currently described in the UFSAR, versus scrubbing radionuclides as described in the LAR). The licensees response also provides historical evidence that operators can initiate drywell spray within the time available. Therefore, the NRC staff concludes that since the environmental conditions affecting operators in the CR following the implementation of this amendment are not significantly different from the conditions assumed to occur in the UFSAR analysis of a LOCA, it is reasonable to assume that any human factors analyses that were previously used to validate these actions should remain valid, and the LAR, therefore, is acceptable.

The revision to TS 3.6.2.6 clarifies the newly credited function of the drywell sprays. Describing the use of the sprays for scrubbing radionuclides in the TS bases provides a useful information reference for the operators. In addition, changes to the TS are assessed and addressed in

operator training programs, and changes to procedures are validated using established procedure modification protocols. These processes provide reasonable assurance that the plant operators will understand the changes.

In addition, the operator actions identified in the LAR are monitored by the Operator Response Time Program at Dresden. This helps to ensure that the newly credited operator actions remain feasible over time.

The considerations described above are consistent with the state-of-the-art human factors principles described in Chapter of 18 of NUREG-0800 and in NUREG-0711/NUREG-1764. Therefore, the NRC staff concludes that since the licensee is adequately applying applicable human factors guidance and is adhering to the intent of GDC-19, with a focus on the safe operation of the plant, (discussed above in Section 2.3.2), the licensees proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on October 5, 2020. The Illinois State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, published in the Federal Register on April 7, 2020 (85 FR 19511), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributors: S. Smith S. Meighan N. Chien B. Green B. FitzPatrick Date of issuance: October 23, 2020

ML20265A240 *via e-mail OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DSS/SCPB/BC* NRR/DRA/ARCB/BC*

NAME RHaskell SRohrer BWittick KHsueh DATE 10/01/2020 10/01/2020 09/24/2020 06/26/2020 OFFICE NRR/DRO/IOLB/BC* NRR/DSS/SNSB/BC* NRR/DSS/STSB/BC* NRR/DNLR/NCSG/BC*

NAME CCowdrey SKrepel VCusumano SBloom DATE 09/03/2020 10/01/2020 08/27/2020 10/01/2020 OFFICE NRR/DEX/EENB/BC* OGC (NLO)* NRR/DORL/LPL3/BC* NRR/DORL/LPL3/PM*

NAME TNavedo STurk NSalgado (RKuntz for) RHaskell DATE 08/28/2020 10/23/2020 10/23/2020 10/23/2020