Text

Application to Increase Technical Specifications Allowable MSIV Leakage Rates, Revise Secondary Containment Surveillance Requirement, and Request Exemption to 10 CFR 50, Appendix J, Option B
	ML19294A304
Person / Time
Site:	Dresden
Issue date:	10/21/2019
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19294A303	List: ML19294A304; RS-19-093, Secondary Containment Drawdown Analysis;
References
	RS-19-093
	Download: ML19294A304 (58)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.90 10 CFR 50.12 RS-19-093 October 21, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Application to Increase Technical Specifications Allowable MSIV Leakage Rates, Revise Secondary Containment Surveillance Requirement, and Request Exemption to 10 CFR 50, Appendix J, Option B

References:

1. Letter from M. Banerjee (U.S. NRC) to C. Crane (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Re:

Adoption of Alternative Source Term Methodology (TAC Nos. MB6530, MB6531, MB6532, MB6533, MC8275, MC8276, MC8277, and MC8278),"

dated September 11, 2006

2. Letter from J. M. Whitman (U.S. NRC) to Technical Specifications Task Force, "Final Safety Evaluation of Technical Specifications Task Force Traveler TSTF-551, Revision 3, 'Revise Secondary Containment Surveillance Requirements' (CAC No. MF5125)," dated September 21, 2017

3. Letter from B. Purnell (U.S. NRC) to B. C. Hanson (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Supplemental Information Needed for Acceptance of License Amendment Request to Revise Technical Specification Requirements for Secondary Containment (EPID L-2017-LLA-0379)," dated January 9, 2018

4. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Withdrawal of Application to Revise Technical Specifications to Adopt TSTF-551, 'Revise Secondary Containment Surveillance Requirements',"

dated January 24, 2018

October 21, 2019 U.S. Nuclear Regulatory Commission Page 2

5. Summary of Pre-application Meeting with Exelon Generation Company, LLC (Exelon) Regarding Forthcoming License Amendment Requests to Revise Technical Specification Requirements for Main Steam Isolation Valve Leak Rate Tests (EPID L-2018-LRM-0088), dated January 2, 2019 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

Several aspects of this request were discussed with the NRC during a December 6, 2018 public pre-submittal meeting (Reference 5).

The proposed amendment alters TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs),"

Surveillance Requirement (SR) 3.6.1.3.10 by revising the combined Main Steam Isolation Valve (MSIV) leakage rate limit for all four steam lines from 86 to 156 standard cubic feet per hour (scfh) for Unit 2 and 218 scfh for Unit 3, respectively. The proposed amendment also revises the leakage rate through each MSIV leakage path from 34 to 62.4 scfh for Unit 2 and 78 scfh for Unit 3, respectively. These proposed changes to the leakage rate limits are based on a revised radiological consequences analysis of the design basis loss of coolant accident (LOCA) in accordance with the DNPS alternative source term (AST) methodology previously approved by the NRC in Reference 1. MSIV leakage will also 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 separate from the station La totals. This change requires an exemption to 10 CFR 50, Appendix J, Option B as described below. A new TS 3.6.2.6, "Drywell Spray" is also added to reflect the crediting of drywell spray for fission product removal as part of the revised LOCA analysis.

In addition, the proposed change revises TS 3.6.4.1, "Secondary Containment," SR 3.6.4.1.1 to address short-duration conditions during which the secondary containment pressure may not meet the SR pressure requirement. The proposed change is consistent with Technical Specifications Task Force Traveler (TSTF) 551, "Revise Secondary Containment Surveillance Requirements," Revision 3 (Reference 2), which was approved by the NRC on September 21, 2017. The proposed change adds a Note to SR 3.6.4.1.1 that allows the secondary containment vacuum limit to not be met for a short duration period provided an analysis demonstrates that one standby gas treatment (SGT) subsystem remains capable of establishing the required secondary containment vacuum. Previous NRC questions raised in Reference 3 regarding the applicability of TSTF-551 to DNPS, which lead to withdrawal of a prior submittal of this change (Reference 4), have now been addressed.

In order to support removal of MSIV leakage from La, DNPS is requesting that the NRC grant an exemption from 1) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A to allow exclusion of the Main Steam Isolation Valve (MSIV) leakage from the overall integrated leakage rate measured when performing a Type A Test, and 2) the requirements of 10 CFR 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.

October 21, 2019 U.S. Nuclear Regulatory Commission Page 3 provides a description and assessment of the proposed changes. The enclosures to Attachment 1 provide the drawdown analysis and LOCA dose consequence analysis that support the assessment of the proposed change. Attachment 2 provides the existing TS pages marked-up to show the proposed TS changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides TS Bases pages marked up to show the associated TS Bases changes and is provided for information only. Attachment 5 provides the proposed exemption to 10 CFR 50, Appendix J, Option B.

The proposed license amendment has been reviewed by the DNPS Plant Operations Review Committee, in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed license amendment by October 21, 2020. The amendment shall be implemented within 60 days following NRC approval.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b}, a copy of this application, with attachments, is being provided to the designated State Officials.

There are no regulatory commitments contained in this submittal. Should you have any questions concerning this submittal, please contact Ms. Rebecca L. Steinman at (630) 657-2831.

I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 21 51 day of October 2019.

Patrick R. Simpson Sr. Manager Licensing Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes

2. Mark-up of DNPS, Units 2 and 3 Technical Specifications Pages

3. Clean DNPS, Units 2 and 3 Technical Specifications Pages

4. Mark-up of DNPS, Units 2 and 3 Technical Specifications Bases Pages - For Information Only

5. Proposed Exemption to Certain 10 CFR 50, Appendix J, Option B Requirements

October 21, 2019 U.S. Nuclear Regulatory Commission Page 4

Enclosures:

A. DRE19-0015, Revision 0a, Dresden Units 2 & 3 Secondary Containment Drawdown Analysis B. DRE05-0048, Revision 5, Dresden Units 2 & 3 Post-LOCA EAB, LPZ, and CR Dose -

AST Analysis cc:

NRC Regional Administrator, Region III NRC Senior Resident Inspector, Dresden Nuclear Power Station NRC Project Manager, Dresden Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Revising MSIV Leakage Rate Limits 2.2 New TS 3.6.2.6 Drywell Spray 2.3 Adoption of TSTF-551 2.4 Mark-ups and Implementation

3.0 TECHNICAL EVALUATION

3.1 Proposed Change to the MSIV Leakage Rate Limits 3.2 New TS 3.6.2.6 Drywell Spray 3.3 Applicability of TSTF-551 Safety Evaluation

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration 4.3 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Changes Page 1 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

The proposed amendment alters TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs),"

Surveillance Requirement (SR) 3.6.1.3.10 by revising the combined Main Steam Isolation Valve (MSIV) leakage rate limit for all four steam lines from 86 to 156 standard cubic feet per hour (scfh) for Unit 2 and 218 scfh for Unit 3, respectively. The proposed amendment also revises the leakage rate through each MSIV leakage path from 34 to 62.4 scfh for Unit 2 and 78 scfh for Unit 3, respectively. These proposed changes to the leakage rate limits are based on a revision of the alternate source term (AST) analysis of the radiological consequences of the design basis Loss of Coolant Accident (LOCA). MSIV leakage will also 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 separate from the station La totals. TS 3.6.2.6, "Drywell Spray," is also added to reflect the crediting of drywell spray for fission product removal as part of the revised LOCA analysis.

In addition, the proposed change revises TS 3.6.4.1, "Secondary Containment," SR 3.6.4.1.1 to address short-duration conditions during which the secondary containment pressure may not meet the SR pressure requirement. The proposed change is consistent with Technical Specifications Task Force Traveler (TSTF) 551 (TSTF-551), "Revise Secondary Containment Surveillance Requirements," Revision 3, which was approved by the NRC on September 21, 2017 (Reference 6.1). The proposed change adds a Note to SR 3.6.4.1.1 that allows the secondary containment vacuum limit to not be met for a short duration period provided an analysis demonstrates that one standby gas treatment (SGT) subsystem remains capable of establishing the required secondary containment vacuum. Previous NRC questions regarding the applicability of TSTF-551 to DNPS in Reference 6.2, which lead to withdrawal of a prior submittal of this change (Reference 6.3), have now been addressed. The portion of TSTF-551 that modifies Standard Technical Specification (STS) SR 3.6.4.1.3 is already incorporated into DNPS, Units 2 and 3 SR 3.6.4.1.2 (Reference 6.13) and is therefore not included in this license amendment request.

2.0 DETAILED DESCRIPTION The Nuclear Regulatory Commission (NRC) approved the use of AST for the evaluation of the onsite and offsite dose consequences for the following Design Basis Accidents (DBA): Loss of Coolant Accident (LOCA), Control Rod Drop Accident (CRDA), Fuel Handling Accident (FHA),

and Main Steam Line Break (MSLB) at DNPS in Reference 6.5. This amendment request incorporates the results of a newly created drawdown analysis (see Enclosure A) and corresponding revision of the LOCA dose consequence analyses (see Enclosure B) to revise the MSIV leakage rate limits in the DNPS, Units 2 and 3 TS, add a TS for drywell spray, and adopt the portion of TSTF-551 that requires a drawdown time.

ATTACHMENT 1 Evaluation of Proposed Changes Page 2 2.1 Revising MSIV Leakage Rate Limits 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 (i.e., inboard) and one outside containment (i.e., outboard). The MSIVs are functionally part of the primary containment boundary and 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.

The MSIV leakage is modeled as an independent leakage pathway in the LOCA radiological consequence evaluation (i.e., separate from the containment leakage pathway). Leakage past the two in-series MSIVs on each MSL is performed in accordance with SR 3.6.1.3.10.

The allowable leakage rate specified in SR 3.6.1.3.10 will be changed to provide unit specific leakage rate limits. 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, of 43.9 psig. As described in plant procedures, the TS leakage rates are calculated using a conversion factor of 1.603 per the extrapolation factor formula for laminar flow from Equation 10.39 of ORNL-NSIC-5 (Reference 6.21) to convert the leakage at the design pressure of 43.9 psig to a TS leakage rate at the test pressure of 25 psig. For Unit 2, the limit will be increased from 34 scfh to 62.4 scfh per MSIV leakage path and the combined limit for all four steam lines will be increased to 156 scfh. Similarly, the Unit 3 allowable leakage rates will be increased to 78 scfh per MSIV leakage path and 218 scfh combined for all four steam lines. The proposed changes reflect a higher, but still conservative allowable leakage rate for the MSIVs based on the results of revised offsite and control room operator dose calculations for the limiting DNPS DBA. The following change to SR 3.6.1.3.10 is proposed. Strikeout indicates proposed deletions and underlined text indicates proposed additions.

Current SR 3.6.1.3.10 Proposed SR 3.6.1.3.10 Verify the leakage rate through each MSIV leakage path is 34 scfh when tested at 25 psig, and the combined leakage rate for all MSIV leakage paths is 86 scfh when tested at 25 psig.

Verify the leakage rate through each MSIV leakage path is 34 62.4 scfh for Unit 2 and 78 scfh for Unit 3 when tested at 25 psig, and the combined leakage rate for all MSIV leakage paths is 86 156 scfh for Unit 2 and 218 scfh for Unit 3 when tested at 25 psig.

The LOCA radiological dose consequence evaluation (Enclosure B) was revised to support the changes to the allowable MSIV leakage limits described above. The revised analysis excludes MSIV leakage from the La totals. This change aligns DNPS with standard industry practice and allows for a more direct comparison when evaluating La performance. TS 5.5.2 and 5.5.12 were evaluated and determined to not be impacted by the increase in allowable MSIV leakage rate or the change to remove MSIV leakage from La. However, removal of MSIV leakage from La has been determined to require an exemption from 10 CFR 50, Appendix J, Option B. The proposed exemption is provided in Attachment 5.

Unplanned MSIV repairs are a significant contributor to increased duration and unplanned radiation exposure during refueling outages. Increasing the MSIV leakage rate limits as shown above will relax operational constraints during outage activities and will improve the overall

ATTACHMENT 1 Evaluation of Proposed Changes Page 3 integrity performance of the MSIVs by reducing the number of maintenance activities. This change will also result in a reduction in personnel exposure due to the decreased valve maintenance being performed on the MSIVs and provide an economic benefit to EGC in terms of direct costs due to the reduction in outage activities. The limiting DNPS design basis accident analysis considering the revised MSIV leakage rate continues to meet the acceptance criteria of 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room (CR) dose consequences.

2.2 New TS 3.6.2.6 Drywell Spray The current licensing basis (CLB) does not credit drywell spray for mitigation of any Updated Final Safety Analysis Report (UFSAR) Chapter 6 or 15 accident analysis. The revised LOCA dose consequence analysis credits the use of drywell sprays for the reduction of airborne activity in the drywell by scrubbing radionuclides from the drywell air space mitigating the consequence of the postulated LOCA event. Because the drywell spray function now meets the requirements of 10 CFR 50.36(c)(2)(ii) Criterion 3 for a system that actuates to mitigate the consequences of a design basis accident (DBA), the requirements for drywell spray are being moved from the Technical Requirements Manual (TRM) to the TS.

TS 3.6.2.6, "Drywell Spray" is added as a new TS because the revised LOCA analysis credits drywell spray for accident mitigation. The proposed LCO requires two drywell spray subsystems to be operable. The TS, including the LCO, applicability, action statements, required completion times, and SRs are patterned after existing drywell spray TS at Edwin Hatch Units 1 and 2, Monticello, Nine Mile Point 2, Peach Bottom Units 2 and 3, and Browns Ferry Units 1, 2, and 3.

The proposed wording for the new DNPS TS 3.6.2.6 is provided below with additional technical rationale provided in Section 3.2.

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 subsystem inoperable.

A.1 Restore drywell spray subsystem to OPERABLE status.

7 days B.

Two drywell spray subsystems inoperable.

B.1 Restore one drywell spray subsystem to OPERABLE 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

ATTACHMENT 1 Evaluation of Proposed Changes Page 4 CONDITION REQUIRED ACTION COMPLETION TIME status.

C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify 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.

In accordance with the Surveillance Frequency Control 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 gas accumulation are sufficiently filled with water.

In accordance with the Surveillance Frequency Control Program 2.3 Adoption of TSTF-551 The safety function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a DBA to ensure the control room operator and offsite doses are within the regulatory limits. The secondary containment requires support systems to maintain the building pressure at less than atmospheric pressure to prevent ground level exfiltration of radioactive material. SR 3.6.4.1.1 requires the secondary containment to be 0.25-inch of vacuum water gauge during normal operation. Following an accident, the SGT System ensures the secondary containment pressure is less than the external atmospheric pressure. 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate 4000 cfm.

ATTACHMENT 1 Evaluation of Proposed Changes Page 5 Secondary containment is a single train system that performs a safety function. Per the guidance in NUREG-1022, Revision 3, "Event Report Guidelines 10 CFR 50.72 and 50.73,"

inoperability of a single train safety system is reportable under 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and 10 CFR 50.73, "Licensee event report system." Currently, failure to meet the secondary containment Limiting Condition for Operation (LCO) or SRs for any period time, even for a brief period much less than the 4-hour Completion Time of TS 3.6.4.1, Condition A, requires declaring secondary containment inoperable and reporting the condition under 10 CFR 50.72 and 10 CFR 50.73.

For the secondary containment to be considered operable, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained by a single operating SGT subsystem. The secondary containment vacuum requirements (which demonstrate leak-tightness) and the SGT System together ensure radioactive material is contained. As long as an SGT subsystem can draw the required vacuum on the secondary containment when needed (as demonstrated by SR 3.6.4.1.3), the secondary containment can perform its safety function.

The proposed change to SR 3.6.4.1.1 will allow a short-duration deviation from SR vacuum acceptance criteria without declaring the secondary containment inoperable, eliminating the attendant reporting requirement provided secondary containment remains capable of meeting its required safety function. Elimination of the reporting requirement when secondary containment remains capable of performing its required safety function during the short duration prior to restoration of vacuum reduces both licensee and NRC resource expenditures.

SR 3.6.4.1.1 requires the secondary containment vacuum to be greater than a required vacuum limit at all times. However, it is possible for the secondary containment vacuum to be momentarily less than the required vacuum of 0.25-inch of water gauge for a number of reasons, such as during wind gusts and during maintenance, testing, or swapping of the normal ventilation subsystems. These conditions do not affect the ability of the SGT System to establish and maintain the required 0.25-inch water gauge vacuum in the secondary containment as assumed in the accident analyses. The following note will be added prior to SR 3.6.4.1.1 to address conditions in which the normal operational secondary containment vacuum is less than the required 0.25-inch water gauge vacuum limit, but one SGT subsystem remains capable of limiting releases from the secondary containment in accordance with the assumptions of the accident analysis.

Current SR 3.6.4.1.1 Revised SR 3.6.4.1.1 Verify secondary containment vacuum is 0.25 inch of vacuum water gauge.

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

Not required to be met for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.

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

ATTACHMENT 1 Evaluation of Proposed Changes Page 6 2.4 Mark-ups and Implementation contains a marked-up version of the DNPS, Units 2 and 3 TS showing the proposed changes described above. Attachment 3 provides the revised (clean) TS pages.

EGC will make supporting change to the TS Bases in accordance with TS 5.5.10, "Technical Specifications (TS) Bases Control Program." Attachment 4 provides the marked-up TS Bases pages. The TS Bases mark-up pages are being submitted for information only. EGC will separately make supporting changes to the TRM to remove TLCO 3.6.a Drywell Spray and update Surveillance Frequency Control Program (SFCP) information; these mark-ups are not included in this document and will be made in accordance with approved plant change processes for these documents.

3.0 TECHNICAL EVALUATION

3.1 Proposed Change to the MSIV Leakage Rate Limits On September 11, 2006 (Reference 6.5) the NRC issued Amendments Nos. 221 and 212 to the Renewed Facility Operating Licenses for DNPS, Units 2 and 3, respectively. These amendments adopt the full implementation of the AST methodology in accordance with 10 CFR 50.67, "Accident Source Term." At that time DNPS elected to retain the TID-14844 assumptions for performing environmental qualification (EQ) analyses.

The basis for the proposed increase in the TS MSIV leakage rate limits is a revision of the RADTRAD 3.03 (Reference 6.7) radiological consequence analysis of the design basis LOCA.

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 revision of the LOCA dose consequence analysis has no impact on the TID-14844 source term used for EQ analysis since it assumes isolation at the inboard MSIV and thus no MSL shine dose in the MSIV room.

The review of the AST license amendment included the assumptions, inputs, and methods used to assess the onsite and offsite /Q values used in the analysis of the postulated DBA dose consequences. The original AST LOCA dose consequence analysis conservatively elected to utilize bounding control room /Q values for both units. The revised LOCA dose consequence analysis uses the originally developed unit-specific control /Q values instead of a single bounding set for both units. Application of the original unit-specific values is a change in input parameter value for Unit 3, but is not called out in Table 3-1 since the numerical /Q values are unchanged from the CLB site-specific values.

A summary of the changes to the methodology and inputs of the revised LOCA analysis compared to the CLB analysis are provided in Table 3-1. Additional information regarding changes to the analysis can be found in Enclosure B, Section 5.8. A discussion of the reason for each change is provided below Table 3-1.

ATTACHMENT 1 Evaluation of Proposed Changes Page 7 Table 3-1: Summary of LOCA Analysis Revisions Design Input Parameter Current Licensing Value Revised Value Additional Information Reduction of MSIV and Containment Leak Rate by 50% at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after LOCA with Respect to RADTRAD Flow Rates Not Credited Credited Enclosure B, Sections 2.1.3

& 4.6.3 Aerosol Deposition in Horizontal MSLs Upstream of Inboard MSIV Not Credited for MSL containing MSIV that failed to close Credited for MSL containing MSIV that failed to close Enclosure B, Section 7.3 Fraction of Containment Leakage that Bypasses the Standby Gas Treatment (SGT) System due to High Winds 0.0 100% (during drawdown period) 0% (following drawdown period)

Enclosure B, Sections 2.1.4

& 5.3.2.8 Percentage of Engineered Safety Feature (ESF) and Containment Leakage that is filtered by the SGT System 100%

0% (during drawdown period) 100% (following drawdown period)

Enclosure B, Section 5.4.10 SGT System Exhaust Charcoal Filter Efficiencies Elemental Iodine:

80%

Organic Iodide:

80%

Elemental Iodine:

90%

Organic Iodide:

90%

Enclosure B, Section 7.9 MSIV Leak Rate Through All Four Lines MSIV Leakage per Line (scfh) 150 scfh @ 43.9 psig 60 60 30 0 (See Figure 2) 250 scfh @ 43.9 psig for Unit 2 and 350 scfh

@ 43.9 psig for Unit 3 Unit 2: 100 100 50 0 (See Figure 1)

Unit 3: 125 125 100 0 Enclosure B, Sections 2.3.3, 4.6.6, & 5.5 (Unit 2) and Appendix A (Unit 3)

Elemental Iodine Removal Efficiency in MSLs 50% Removal Efficiency Credited for Accident Duration Time Dependent Removal Efficiency Enclosure B, Section 2.3.2 Control Room (CR)

Unfiltered Inleakage during Normal Operation 60,000 cfm (includes ingress/egress inleakage of 10 cfm) 4000 cfm (includes ingress/egress inleakage of 10 cfm)

Enclosure B, Sections 4.8.6

& 5.6.6 Particulate (Aerosol)

Deposition/Plateout Model in Containment Powers 10 percentile Model Not credited Enclosure B, Sections 2.1.3, 4.6.2, and 5.8 Elemental Iodine Removal in Containment via Natural Deposition Credited Not credited Enclosure B, Sections 2.1.3, 4.6.5, & 5.8 Total Containment Leakage 3% volume/day minus MSIV leakage 3% volume/day Enclosure B, Section 5.8

ATTACHMENT 1 Evaluation of Proposed Changes Page 8 Table 3-1: Summary of LOCA Analysis Revisions Design Input Parameter Current Licensing Value Revised Value Additional Information Drywell Spray Not credited Credited Enclosure B, Sections 2.1.3, 4.6.7, & 7.11 Core Inventory Bounds existing fuel types Bounds expected future fuel types Enclosure B, Section 2.1.1 Reduction of MSIV and Containment Leak Rate by 50% at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after LOCA with Respect to RADTRAD Flow Rates Primary containment and the MSIVs are assumed to leak at the peak pressure leak rate for the initial 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the LOCA. Reduction in the containment leakage and MSIV leakage after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the maximum leakage is credited in the revised dose consequence analysis.

In accordance with Darcys equation for compressible fluid flow through orifices, the volumetric flow rate is proportional to the square root of the driving pressure and inversely proportional to the square root of the fluid density. Using this relationship, a pressure reduction of 75% leads to a flow rate reduction of 50%. Since the flow rates are based on a maximum drywell pressure of 43.9 psig, pressures less than approximately 11 psig will result in a reduction in flow of at least 50%. The containment system response analysis shows that the drywell pressure is 21.8 psia (7.1 psig) at 40,000 seconds (~11 hours) following a LOCA, well below the maximum drywell driving pressure of 43.9 psig. Additionally, the maximum drywell temperature occurs late in the event but since the calculated flow rates are already based on the highest temperature (lowest density) no credit is taken for the effect of reduced volumetric flow due to increased density.

Therefore, assuming a leak rate reduction of 50% of the maximum at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the event is conservative with respect to the actual expected leakage rate.

The aerosol deposition is modelled using a filter efficiency in RADTRAD. As outlined in AEB-98-03, this approach is developed based on an assumption that the aerosol concentration is constant in a well-mixed volume. A changing aerosol concentration could potentially cause an inaccurate filter efficiency. To address this concern, the aerosol removal efficiencies in RADTRAD are conservatively calculated based on the non-reduced flow rates (i.e., the removal efficiencies do not credit the 50% reduced flow rate at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Aerosol Deposition in Horizontal Main Steam Lines Upstream of Inboard MSIV The DNPS main steam piping from the reactor pressure vessel (RPV) to the outboard MSIV is ASME Class 1 seismically analyzed to assure the piping wall integrity during and after a safe shutdown earthquake (SSE) event. Based on this structural integrity, the four MSL piping sections between the RPV nozzle and outboard MSIVs are assumed to remain intact providing horizontal pipe surface area and volume for aerosol deposition.

The CLB analysis assumes that the horizontal MSL volume upstream of the failed inboard MSIV does not remove aerosols and only credits removal in portions of the MSL piping upstream of two intact inboard MSIVs. This assumption is based on a postulated main steam line pipe break just upstream of a MSIV. The initiating event is a large pipe break of a recirculation suction line

ATTACHMENT 1 Evaluation of Proposed Changes Page 9 with a failure of the inboard MSIV of a steam line to close. Multiple simultaneous pipe breaks are not considered as part of the design basis containment analysis; as a result, the LOCA dose analysis does not consider multiple simultaneous pipe breaks.

The CLB aerosol removal methodology does not consider revaporization of aerosols that have plated-out in the MSLs. This model is not changed in the revised LOCA dose consequence analysis. Di Lemma et al (Reference 6.23, Section 4.2) suggests a revaporization threshold of

>1,100ºK (>1,456ºF) for Cesium Iodide (CsI) aerosols. Temperatures in the steam lines are expected to remain below 600ºK per Figure 7 of Reference 6.9. Based on the starting temperature of the steam lines and the expected temperature rise due to the LOCA, it is not expected that temperatures would reach levels such that significant aerosols would become revaporized in the MSLs.

Fraction of Containment Leakage that Bypasses the Standby Gas Treatment (SGT) System due to High Winds Leakage from the primary containment will collect in the free volume of the secondary containment and be released to the environment via the ventilation system exhaust. Following a LOCA, the SGT System fans start and drawdown the secondary containment to create a negative pressure with reference to the environment. Once a negative pressure is established, the pressure differential ensures that leakage from the primary containment is collected and processed by the SGT System. The SGT System exhaust is processed through charcoal and HEPA filter media before release to the environment.

A new analysis to determine the time duration to establish a sustained negative pressure in the secondary containment has been performed (See Enclosure A). This analysis assumes following an accident, isolation of the secondary containment ventilation system, failure of the primary SGT System fan to start, and auto start of the standby SGT System fan. The results of this analysis determined that the secondary containment drawdown is achieved in 22 minutes.

For conservatism, the revised LOCA dose consequence analysis assumes a 25-minute drawdown period. In accordance with RG 1.183, Appendix A, Section 4.2, during this 25-minute drawdown period, leakage from the primary containment into the secondary containment is assumed to be released directly to the environment as a ground-level release.

Percentage of Engineered Safety Feature (ESF) and Containment Leakage that is Filtered by the SGT System During the initial 25-minute drawdown period needed to establish the TS required vacuum, the revised LOCA analysis assumes that the ESF and containment leakage is an unfiltered release from the secondary containment to the environment. After the required vacuum has been established, the revised analysis assumes that 100% of ESF and containment leakage is filtered by the SGT System. This is consistent with RG 1.183, Appendix A, Section 4.2.

SGT System Exhaust Charcoal Filter Efficiencies Per Reference 6.10, TS 5.5.7.c, the laboratory testing methyl iodide penetration acceptance criteria for the SGT System vent charcoal filter is less than 2.5%. Reference 6.14 requires a safety factor of at least 2 be used to determine the filter efficiencies credited in the design basis

ATTACHMENT 1 Evaluation of Proposed Changes Page 10 accident dose consequence analysis, as shown below.

Testing methyl iodide penetration (%) = (100% - )/safety factor = (100% - )/2 Where = SGT System Vent Charcoal Filter efficiency to be credited in the analysis SGT System Vent Charcoal Filter 2.5% = (100% - )/2 5% = (100% - )

= 100% - 5% = 95%

Conservatively, an SGT System vent charcoal filter efficiency of 90% is credited in the analysis.

MSIV Leakage The four MSLs, which penetrate the primary containment, are automatically isolated by the MSIVs in the event of a LOCA. There are two MSIVs on each steam line, one inside containment and one outside containment. The MSIVs are functionally part of the primary containment boundary and design leakage through these valves provides a leakage path for fission products to bypass the secondary containment and enter the environment as a ground-level release. 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 control room ground release atmospheric dispersion coefficients are lower than the Unit 2 values. All other inputs are common to both Units 2 and 3.

The radiological consequences from postulated MSIV leakage are analyzed and combined with the radiological consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.

All four MSL piping sections between the RPV nozzle and outboard MSIVs used in the MSIV leakage release paths remain intact and can perform their safety function during and following an 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 is assumed to occur as follows:

1) 100 scfh for Unit 2 and 125 scfh for Unit 3 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 for Unit 2 and 125 scfh for Unit 3. 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.

ATTACHMENT 1 Evaluation of Proposed Changes Page 11

3) 50 scfh through second intact steam line for Unit 2 and 100 scfh for Unit 3. 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.

Figure 1 shows the MSIV leakage model credited in the revised analysis while Figure 2 shows the MSIV leakage model credited in the current licensing basis analysis. The leakage values in these figures correspond to the Unit 2 model during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a LOCA. The flow distributions are chosen to be consistent with the CLB analysis, which conservatively ignores aerosol and elemental removal in the fourth steam line.

Time Dependent Elemental Iodine Removal Efficiency in MSLs Gaseous iodine tends to deposit on the piping surface by chemical adsorption. Elemental iodine, being the most reactive, has the highest deposition rate. The iodine deposited on the surface undergoes both physical and chemical changes and can be re-emitted as an airborne gas (re-suspension) or permanently fixed to the surface (fixation). RG 1.183, Appendix A, Section 6.5 (Reference 6.6) indicates that the methodology given in Reference 6.9 provides acceptable models for deposition of iodine on pipe surfaces. This methodology is used to determine the deposition and resuspension rates of elemental iodine in the revised analysis.

The CLB analysis assumes 50% removal efficiency for the duration of the event, but the revised model leads to modeled removal efficiency rates below 50% for all pathways for time periods less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The time steps used are 2, 8, 24, 48, 72, 96, 240, and 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months .

Sensitivity runs show that modeling time dependent elemental iodine removal is conservative because it leads to less overall removal than the constant 50% removal efficiency modeled in the CLB analysis.

Control Room (CR) Unfiltered Inleakage during Normal Operation The revised CR unfiltered inleakage during normal operation has been reduced from 60,000 cubic feet per minute (cfm) to 4,000 cfm. The normal outside intake flow rate is 2,000 cfm +/- 10% so assuming the inleakage is double the nominal intake is conservative for this analysis. This inleakage rate bounds the latest tracer gas test inleakage of 534 cfm (495 +/- 39 cfm) during isolation - recirculation mode. The tracer gas test was last performed in February 2015, in accordance with TS 5.5.13, "Control Room Envelope Habitability Program."

This tracer gas test data corresponds to a toxic gas scenario where recirculation in the control room is being provided by Control Room Emergency Ventilation System (CREVS) train B. This scenario does not represent normal operation of the CR HVAC with 2,000 cfm of intake but demonstrates that even with a negative pressure in the CR with no makeup, an inleakage value of 534 cfm is bounded by the modeled 4,000 cfm.

The CR unfiltered inleakage during emergency operation is unchanged from the previously approved value of 395 cfm. This value continues to be supported by ongoing tracer gas testing representing emergency operation following a LOCA.

Page 12 Figure 1: Revised MSIV Leakage RADTRAD Nodalization DRYWELL Turbine Building No Mixing E N V I R O N M E N T Path 3 -

Drywell to Intact MSL 1 (100 scfh)

Path 4 - Drywell to Intact MSL 1 (100 scfh)

- includes RPV nozzle to inboard MSIV removal Path 7 - Drywell to Intact MSL 2 (50 scfh)

- includes RPV nozzle to inboard MSIV removal Path 8 - Intact MSL 2 to environment (50 scfh)

- includes inboard MSIV to outboard MSIV removal Path 6 -

Drywell to Intact MSL 2 (50 scfh)

Path 5 - Intact MSL 1 to environment (100 scfh)

- includes inboard MSIV to outboard MSIV removal Path 2 - MSL 1 With MSIV Failure to environment (100 scfh) - includes RPV nozzle to outboard MSIV removal Path 1 -

Drywell to MSL 1 With MSIV Failure (100 scfh)

Compartment 2 Failed MSL 1 RPV Nozzle to outboard MSIV Compartment 3 Intact MSL 1 RPV Nozzle to inboard MSIV Compartment 4 Intact MSL 1 RPV inboard MSIV to outboard MSIV Compartment 5 Intact MSL 2 RPV Nozzle to inboard MSIV Compartment 6 Intact MSL 2 inboard MSIV to outboard MSIV

Page 13 Figure 2: Current Licensing Basis MSIV Leakage RADTRAD Nodalization DRYWELL Turbine Building No Mixing E N V I R O N M E N T Path 3 -

Drywell to Intact MSL 1 (60 scfh)

Path 4 - Drywell to Intact MSL 1 (60 scfh)

- includes RPV nozzle to inboard MSIV removal Path 7 - Drywell to Intact MSL 2 (30 scfh)

- includes RPV nozzle to inboard MSIV removal Path 8 - Intact MSL 2 to environment (30 scfh) - includes inboard MSIV to outboard MSIV removal Path 6 -

Drywell to Intact MSL 2 (30 scfh)

Path 5 - Intact MSL 1 to environment (60 scfh)

- includes inboard MSIV to outboard MSIV removal Path 2 - MSL 1 With MSIV Failure to environment (60 scfh) - includes inboard MSIV to outboard MSIV removal Path 1 - Drywell to MSL 1 With MSIV Failure (60 scfh)

Compartment 2 Failed MSL 1 RPV Nozzle to outboard MSIV Compartment 3 Intact MSL 1 RPV Nozzle to inboard MSIV Compartment 4 Intact MSL 1 RPV inboard MSIV to outboard MSIV Compartment 5 Intact MSL 2 RPV Nozzle to inboard MSIV Compartment 6 Intact MSL 2 inboard MSIV to outboard MSIV

ATTACHMENT 1 Evaluation of Proposed Changes Page 14 The CLB analysis assumes 60,000 cfm as an unfiltered inleakage rate based on a sensitivity analysis performed to show which flow rate would lead to an approximately equilibrium activity between the environment and the control room (Reference 6.4). However, this large flow rate would over-pressurize the control room and is considered excessively conservative for this reason. The RADTRAD model includes 6,200 cfm of total flow (4,000 cfm inleakage and 2,200 cfm intake) into the control room during the first 40 minutes following the LOCA when the normal control room ventilation system is assumed to be operating.

Particulate (Aerosol) Deposition/Plateout Model in Containment The revised analysis removes credit for the Powers deposition model in the drywell. There is approximately 32,250 ft2 of surface area available in the drywell for deposition and plateout of aerosols. However, none of this area is credited in the RADTRAD model. The only surface area credited for deposition and plateout is upstream of the outboard MSIVs in three of the four MSLs. This surface area is 237 ft2.

Elemental Iodine Removal in Containment via Natural Deposition RADTRAD Error Notice No. 18 indicates that sprays and natural deposition should not be modeled concurrently unless a combined decontamination factor (DF) is calculated. Credit for drywell sprays to remove elemental iodine in containment is included in the revised LOCA dose consequence analysis. In lieu of modeling these two removal mechanisms separately or calculating a combined decontamination factor, credit for natural deposition is conservatively removed entirely. This change conservatively prolongs the time until the 200 DF limit is reached.

Total Containment Leakage The CLB analysis reduced the total containment leakage rate by subtracting out the MSIV leakage rate so that the modeled containment leakage rate is less than 3 volume percent per day. This was modeled in the CLB analysis by assuming that the containment leakage exits containment to a "void" region in the MSIV leakage model and the MSIV leakage exits containment to a "void" region in the containment leakage volume. The revised analysis assumes that the containment is leaking at its full design rate of 3 volume percent per day and conservatively does not consider leakage escaping to a void region (where it would not contribute to dose). The Appendix J testing program will be updated to reflect separating MSIV leakage from the overall containment leakage pending approval of the exemption request in.

Drywell Spray RG 1.183, Appendix A, Section 3.3, allows the licensees to take a 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 of the Standard Review Plan (SRP)

(Reference 6.15). DNPS TRM 3.6.a requires (Reference 6.11), "Two drywell spray subsystems shall be operable." Since the new LOCA analysis credits spray for consequence mitigation, this license amendment request moves these TRM requirements to the Technical Specifications.

(See Sections 2.2 and 3.2 for additional details regarding the drywell spray TS.)

ATTACHMENT 1 Evaluation of Proposed Changes Page 15 DNPS UFSAR Section 6.2.2 states that containment cooling is the operating mode of the Low Pressure Coolant Injection (LPCI) subsystem that provides three different containment cooling functions: suppression pool cooling, drywell spray, and suppression chamber spray. The term containment spray, as used within the UFSAR, refers to drywell spray and suppression chamber spray collectively. Containment cooling functions are manually initiated. Equipment and piping in the LPCI system that feeds the containment spray nozzles are safety-related. Even though using the drywell spray subsystem for scrubbing radionuclides from the drywell air space was not considered a safety-related function as part of the original design basis, the operational status of the system is verified through existing TRM surveillances that are being moved to the TS. (See Section 3.2 for additional details.) Additionally, drywell spray is currently modeled in the containment response accident analysis for removing drywell heat to lower drywell temperature and pressures after the peak containment pressure occurs (i.e., drywell spray is not credited for peak pressure mitigation).

Drywell sprays are assumed to start 10 minutes following event initiation and continue for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Operators are directed to start drywell sprays when drywell pressure is above 9 psig and the drywell pressure and temperatures are below the drywell spray initiation limit curve.

Both criteria are met before 10 minutes following a LOCA. Reasonable assurance of the timeliness of this manual action is provided by the Operator Response Time Program. The Operator Response Time Program contains the list of operator actions that are performed within a specified time, which are credited in the plants design and licensing basis. This program directs the operator to manually initiate containment spray (which includes drywell and/or suppression pool spray) for a design basis LOCA at 10 minutes after the event.

Per the containment pressure response analysis, the drywell remains above 7 psig until at least 40,000 seconds (11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months ) following the event, so even though the sprays can continue as long as containment pressure is above 6 psig (the pressure needed to maintain adequate NPSH for the LPCI pumps), drywell sprays are assumed to stop at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to conservatively bound the anticipated operator action to continue sprays until drywell pressure reaches atmospheric conditions.

In accordance with RG 1.183, Appendix A, Section 3.3, the maximum decontamination factor for elemental iodine is based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate divided by the activity of iodine remaining at some time after decontamination. Also, the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The revised analysis determines that the elemental iodine reaches a DF of 200 at 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and aerosol iodine mass reaches a DF of 50 at 2.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

After 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months the elemental iodine removal via spray is terminated and after 2.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the aerosol removal coefficient is reduced by a factor of 10 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months post-accident when the drywell spray is assumed to be terminated.

A comparison between the SRP Section 6.5.2 review items for drywell spray and the discussion of how each item is addressed is provided in Section 2.1.3 of Enclosure B.

ATTACHMENT 1 Evaluation of Proposed Changes Page 16 Core Inventory A new bounding core inventory based on a combination of increased core average exposure (CAVEX) to 43 GWd/MTU and an enrichment range between 3.7 wt% U-235 and 4.5 wt%

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 Table 3-2 of this LAR are based on this bounding CAVEX source term.

Overall Conservatisms The regulatory requirement is that the limiting DNPS DBA meet the acceptance criteria of 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room dose consequences. The current licensing basis dose consequence model for DNPS, as well as the proposed changes to this model are consistent with current NRC guidance in RG 1.183 Revision 0. The results of the revised analysis conclude that the dose consequences remain below the relevant acceptance criteria.

As described above, each of the changes made to the LOCA dose consequence analysis listed in Table 3-1 are individually evaluated with respect to the technical basis for the change and how the individual change impacts the analysis. To ensure the analysis as a whole remains conservative and continues to meet the relevant acceptance criteria, a holistic review of the model changes, as well as the decisions to not change certain inputs and models, was performed to confirm no inadvertent non-conservatisms have been introduced.

The most significant change in the AST methodology that allows for the larger MSIV leak rates requested in this amendment is the crediting of drywell spray as an effective means of removing airborne radioactivity in containment and thus mitigating the overall dose consequences of a design basis LOCA. RG 1.183 Appendix A Section 3.3 is prescriptive with respect to crediting drywell spray removal and this license amendment request demonstrates compliance with this guidance.

RG 1.183, Appendix A, Section 4.5 states that deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis. RIS 2006-04 (Reference 6.16) reiterates that Accident Evaluation Branch (AEB) Report, AEB-98-03 (Reference 6.8), methods are plant specific and require appropriate justification for application to other plant designs. The RIS goes on to discuss specific NRC concerns associated with the effects of aerosol particle size distribution and settling velocity on deposition. The original DNPS SE associated with approval of the AST methodology (Reference 6.5) approved the use of AEB-98-03 deposition models with reference to specific assumptions that ensured reasonable assurance of conservatism in the overall estimated dose consequences. Examples of conservatism cited in the AST SE regarding the acceptability of AEB-98-03 are as follows:

Not crediting a reduction in drywell pressure or MSIV leak rate after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (still true, since reduction is not used for the calculation of aerosol removal efficiencies),

Utilizing a more conservative settling velocity (40th percentile) than the median (unchanged), and

ATTACHMENT 1 Evaluation of Proposed Changes Page 17 Assuming a 50% elemental iodine removal efficiency (new time dependent model has lower overall removal than a constant 50% removal efficiency)

The methodology associated with modeling aerosol removal within the MSLs is unchanged or conservative with respect to the current DNPS licensing basis approved except for two changes:

(1) the revised analysis replaces removal of containment aerosols due to natural deposition with removal due to drywell spray and (2) the revised analysis credits a small additional area of piping upstream of the MSIV for deposition. Both of these changes were made in a manner that consistent with the current guidance in RG 1.183. Understanding the underlying physical phenomena of each removal mechanism is critical to determining the impact of crediting spray and deposition removal methods concurrently.

NUREG/CR-0009 (Reference 6.18) is a compilation of experimental and theoretical information used by the NRC to develop the accident analysis spray removal methodology. This report is primarily based on the containment systems experimental data described in report BNWL-1457 (Reference 6.19). Per NUREG/CR-0009, aerosol removal by containment sprays is primarily due to the following mechanisms:

Brownian diffusion Diffusiophoresis Interception Inertial impaction In addition, NUREG/CR-0009 states that deposition of particles on wall surfaces (either containment walls or MSL pipe walls) is due to the following mechanisms:

Diffusion Thermophoresis Diffusiophoresis Turbulence in the wall boundary layer The spray removal coefficients used in the revised LOCA dose consequence analysis are based on the conservative values in Section 6.5.2 of NUREG-0800 (Reference 6.15). The values assume that the ratio of a dimensionless collection efficiency to the average spray drop diameter should be 10 per meter initially (i.e., 1% efficiency for spray drops of 1 millimeter in diameter) and change abruptly to 1 spray drop per meter after the aerosol mass has been depleted by a factor of 50 (i.e., 98% of the suspended mass is 10 times more readily removed than the remaining 2%). Section J3.2.2 of NUREG-75/014 (Reference 6.20) provides the technical basis for the formula used Section 6.5.2 of NUREG-0800 and in the revised analysis.

NUREG-75/014 Section J3.2.2 also provides the correlation to determine spray lambdas. The spray lambda calculation assumes that diffusiophoresis is not a mechanism for spray removal.

This is confirmed by Figure VII J-4 of NUREG-75/014.

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

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

ATTACHMENT 1 Evaluation of Proposed Changes Page 18 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.

Summary The revised LOCA dose consequence analysis performed in support of this license amendment includes the changes described in Table 3-1 above. The results of the revised LOCA analysis indicate that the total post-LOCA Exclusion Area Boundary (EAB), Low Population Zone (LPZ),

and CR doses are within the regulatory limits for Total Effective Dose Equivalent (TEDE)

(Table 3-2). Enclosure B separately models Unit 2 and 3. The methodology between the units is identical except for the MSIV leakage limit assumptions (a total design leak rate of 250 scfh for Unit 2 and 350 scfh for Unit 3) and the unit-specific control room /Q values. Table 3-2 provides the bounding dose results (i.e., Unit 2 and 3 results are mixed together). The control room dose reported in the table corresponds to the case evaluating a design leakage of 250 scfh from Unit 2. The EAB and LPZ doses correspond to the case evaluating a design leakage of 350 scfh from Unit 3. The control room dose from a release from Unit 3 are lower than the reported Unit 2 control room dose due to crediting the lower site-specific Unit 3 /Q value. Results from each of the cases can be found in Enclosure B.

Table 3-2: LOCA Dose Consequence Summary Post-LOCA Activity Release Path Post-LOCA TEDE Dose (Rem)

Receptor Location Control Room EAB LPZ 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 3.57E+00 8.79E-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 4.86E+00 3.66E+00 1.24E+00 CLB Doses 4.84E+00 1.64E+00 7.81E-01 Allowable TEDE Limit 5.00E+00 2.50E+01 2.50E+01 EQ doses are not impacted by the changes associated with this LAR because the current EQ design basis does not include source term in the main steam lines downstream of the MSIVs.

For the Technical Support Center (TSC) and other areas requiring plant personnel access, assessments confirmed that the radiation exposures would remain within regulatory limits with no new operator actions required.

ATTACHMENT 1 Evaluation of Proposed Changes Page 19 3.2 New TS 3.6.2.6 Drywell Spray The proposed LCO requires two drywell spray subsystems to be operable in MODES 1, 2, and

3. Each of the two drywell spray subsystems contain two pumps, one heat exchanger, drywell spray valves, and a spray header inside the drywell. The proposed SRs are necessary to ensure that the drywell spray flow path is clear such that the flow rate of 2,352 gpm used to calculate the spray removal coefficient in the LOCA dose technical evaluation, DRE05-0048 (Enclosure B), remains valid. The conservative drywell spray flow rate of 2,352 gpm is based on each of the 160 drywell spray nozzles providing 14.7 gpm.

Per UFSAR Section 6.2.1.3.3 the design basis drywell spray flow is 4,750 gpm and wetwell spray flow rate is 250 gpm. Existing SR 3.6.2.3.2 requires that each required LPCI pump develops a flow rate greater than or equal to 5000 gpm while operating in the suppression pool cooling mode, which is substantially greater than the 2,352 gpm assumed for the spray removal coefficient in DRE05-0048.

SR 3.6.2.6.1 ensures that an available flow path exists between the LPCI pumps and the drywell spray nozzles following a LOCA. SR 3.6.2.6.2 ensures that there are no blockages that would affect the spray pattern which would invalidate the radionuclide removal coefficients calculated in DRE05-0048. SR 3.6.2.6.3 ensures that the normally water-filled lines of the drywell spray subsystem do not have gas accumulation to prevent water hammer damage to LPCI components.

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:

ATTACHMENT 1 Evaluation of Proposed Changes Page 20 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 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.

3.3 Applicability of TSTF-551 Safety Evaluation Prior to this license amendment request, the DNPS, Units 2 and 3 CLB LOCA radiological dose consequence analysis assumed that the secondary containment pressure is below atmospheric pressure coincident with the time at which the LOCA event occurs. It also assumed the SGT automatically starts and maintains the negative pressure such that no exfiltration occurs during the event. As a result of the CLB calculation assumptions, TSTF-551 did not apply.

A new plant-specific calculation was performed to determine the reactor building drawdown time. The drawdown calculation was performed using GOTHIC 8.2 to calculate the secondary containment temperature and pressure response. GOTHIC error notices pertaining to version 8.2 were reviewed and none were identified which are applicable to the model used in the drawdown analysis. The following RG 1.183 (Reference 6.6), Section 4.3 assumptions were included in the drawdown calculation:

1. The outside air temperature is assumed to remain constant at the summer design temperature of 93F during summer conditions and at -6F during winter conditions (Reference 6.12, Section 9.4.7). Per RG 1.183 the ambient temperature should be the 1-hour average value that is exceeded only 5% (for summer conditions) and 95% (for winter conditions) of the total number of hours in the data set. The assumed temperatures reflect the summer 1% exceedance value of 91F and conservatively bound the winter 99% exceedance value of 0F for the Dresden area from Reference 6.17, Section 2.3.2.

ATTACHMENT 1 Evaluation of Proposed Changes Page 21

2. A maximum wind speed of 24 mph was assumed for the analysis. Per RG 1.183, the wind speed to be assumed is the 1-hour average value that is exceeded only 5% of the total number of hours in the data set. A wind speed of 24 mph is exceeded less than 5%

of the time at elevations of 35 ft and 150 ft, which bound the height of the Reactor Building. The assumed wind speed is conservative compared to the 5% wind speed of 19 mph for West Chicago, Illinois from Reference 6.17, Table 1A.

Four different cases were analyzed with the GOTHIC model to envelop the assumed environmental conditions: 1) summer with no wind, 2) summer with wind, 3) winter with no wind, and 4) winter with wind. The following sequence of major events are postulated for this calculation:

1. The units are both in normal operation with the secondary containment ventilation system maintaining the specified secondary containment vacuum pressure.

2. A design basis LOCA is assumed to occur with a concurrent loss of offsite power (LOOP).

3. The secondary containment is isolated by closing the secondary containment ventilation system isolation valves and the ventilation system fans are tripped.

4. The diesel generators start and the SGT System fans are loaded onto the diesel generator (DG) buses at 13 seconds after the LOCA occurs.

5. The primary SGT System fan fails to start after being loaded onto the DG bus and the standby SGT System fan starts and the isolation valves begin to open after a 20 second time delay.

6. The SGT System flow rate is controlled to 3975 cfm after the SGT System isolation valves open. 3975 cfm corresponds to a nominal SGT subsystem flowrate of 4000 cfm with a reduction of 25 cfm to account for potentially dirty filters.

The results of this drawdown calculation determined that the drawdown time is longer for the cases with wind (Cases 2 and 4) due to the lower outside air pressures on the downwind side of the secondary containment. The winter cases (Cases 3 and 4) have shorter drawdown times than the summer cases (Cases 1 and 2). The limiting drawdown time for the summer conditions with wind (Case 2) is 1,334 seconds. Therefore, the design basis secondary containment drawdown time is 1,334 seconds, or approximately 22 minutes.

The revised LOCA radiological dose consequence analysis considers a secondary containment positive pressure period of 25 minutes. The bounding results of the revised radiological consequences (see Table 3-2 above) do not exceed the TEDE limits.

EGC has confirmed for DNPS that the brief, inadvertent, simultaneous opening of both an inner and outer personnel access door during normal entry and exit conditions, and their prompt closure by normal means, is bounded by the revised radiological dose consequence analysis.

In the unlikely event that an accident would occur when both personnel access doors are open for entry or exit, the brief time required to close one of the doors is small compared to the

ATTACHMENT 1 Evaluation of Proposed Changes Page 22 25 minute positive pressure period assumed in the accident analysis for reducing the post-accident secondary containment pressure to -0.25-inch of vacuum water gauge and will not result in an increase in any onsite or offsite dose.

Considering the new drawdown analysis and revised LOCA dose consequence results, EGC has determined that TSTF-551 is now applicable to DNPS. EGC has reviewed the Safety Evaluation for TSTF-551 provided to the Technical Specifications Task Force in a letter dated September 21, 2017 (Reference 6.1). This review included a review of the NRC evaluation, as well as the information provided in TSTF-551. EGC has concluded that the justifications presented in TSTF-551 and the Safety Evaluation prepared by the NRC are applicable to DNPS, Units 2 and 3 and justify this amendment for the incorporation of the changes to the DNPS, Units 2 and 3 TS.

The DNPS Units 2 and 3 SR 3.6.4.1.2 already contains the modification acknowledging that secondary containment access openings may be open for entry and exit. Therefore, the proposed change does not contain this portion of TSTF-551.

The Traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). DNPS, Units 2 and 3 were not licensed to the 10 CFR 50, Appendix A, GDC. The DNPS, Units 2 and 3 Updated Final Safety Analysis Report (UFSAR), Section 3.1, "Conformance with NRC General Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concluded that the plant specific requirements are sufficiently similar to the Appendix A GDC.

This difference does not alter the conclusion that the proposed change is applicable to DNPS, Units 2 and 3.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(2), "Limiting conditions for operation," subsection (ii) provides four criteria for which the Technical Specifications (TS) must contain a limiting condition for operation (LCO).

Criterion 3 requires an LCO for 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. The revised LOCA dose consequence analysis credits the use of drywell sprays for the reduction of airborne activity in the drywell by scrubbing radionuclides from the drywell air space mitigating the consequence of the postulated LOCA event. As a result, a new TS 3.6.2.6, "Drywell Spray" is proposed.

10 CFR 50.36(c)(3), "Surveillance requirements," states that SRs are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation are met. The proposed changes to SR 3.6.1.3.10 continue to ensure that leakage through the bypass leakage pathways and the main steam lines is maintained within the values assumed in the LOCA radiological consequences analysis, and, therefore, the LCO will be met.

The proposed change to SR 3.6.4.1 does not alter the design of secondary containment or its ability to establish an essentially leak-tight barrier against the uncontrolled release of

ATTACHMENT 1 Evaluation of Proposed Changes Page 23 radioactivity, therefore the limiting conditions for operation will be met. The new drywell spray TS includes three SRs to ensure a spray flow path capable of delivering the flow rate used to calculate the spray removal coefficient in the LOCA radiological consequences analysis.

10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," identifies requirements for establishing a program for qualifying electric equipment that is important to safety as defined in 10 CFR 50.49(b). The revision of the loss-of-coolant (LOCA) dose consequence analysis has no impact on the TID-14844 source term used for equipment qualification analysis since it assumes isolation at the inboard MSIV and thus no MSL shine dose in the MSIV room. As a result, no new equipment needs to be added to the EQ Program and no existing EQ zone classifications change as a result of the revised LOCA dose consequence analysis.

10 CFR 50.67, "Accident source term," establishes acceptable radiation dose limits resulting from design basis accidents for an individual located at the exclusion area boundary or low population zone, and for occupants of the control room. The analyses performed for DNPS demonstrate that the calculated radiological consequences of a design basis LOCA with increased leakage through the MSLs meet the radiation dose limits specified in 10 CFR 50.67.

10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," describes test requirements that provide assurance that the primary containment, including those systems and components that penetrate the primary containment, do not exceed the allowable leakage rate values specified in the TS and their associated bases.

The proposed amendment maintains compliance with the requirements of 10 CFR 50, Appendix J, Option B when consideration of the exemption request for removing main steam isolation valve leakage from La is included.

Regulatory Guide (RG) 1.183, dated July 2000, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides guidance for implementation of 10 CFR 50.67, including assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an alternate source term (AST).

Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," provides guidance to ensure that the appropriate level of technical detail is considered in AST analyses and included in AST submittals.

Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," describes the review plan for amendments implementing AST for radiological consequences.

The revision to the LOCA radiological consequence analysis is performed using the AST methodology which has been established as the licensing basis for this accident. The regulatory requirements provided in 10 CFR 50.67 and guidance in RG 1.183 and Standard Review Plan 15.0.1 are used in the revised analysis. The RADTRAD 3.03 computer code used to perform the revision of the LOCA analysis has been accepted by the Nuclear Regulatory Commission for use in radiological dose analyses. The calculated Total Effective Dose Equivalent (TEDE) doses to the Control Room, Exclusion Area Boundary (EAB), and to the Low

ATTACHMENT 1 Evaluation of Proposed Changes Page 24 Population Zone (LPZ) are all below the regulatory dose limits.

Based on the considerations discussed above, it is concluded that, (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operating in the proposed manner, (2) activities will be conducted in compliance with NRC regulations, and (3) the approval and issuance of this proposed amendment will not be inimical to the common defense and security of the health and safety of the public.

4.2 No Significant Hazards Consideration Overview Exelon Generation Company, LLC (EGC) requests an amendment to revise Technical Specifications (TS) Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)," by revising Surveillance Requirement (SR) 3.6.1.3.10 for main steam isolation valve (MSIV) leakage rates.

The proposed amendment would increase the allowable leakage rate through each MSIV leakage path and the combined leakage rate limit for all four steam lines. EGC also requests the addition of TS 3.6.2.6, "Drywell Spray." The revised LOCA analysis credits the use of drywell sprays for the reduction of airborne activity in the drywell by scrubbing radionuclides from the drywell air space. Because the drywell spray function now meets the requirements of 10 CFR 50.36, the surveillance requirements are being moved from the technical requirements manual to the technical specifications.

In addition to the above changes for MSIV leakage rate, EGC requests adoption of a portion of TSTF-551, "Revise Secondary Containment Surveillance Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Dresden Nuclear Power Station (DNPS), Units 2 and 3 Technical Specifications (TS). The proposed change revises SR 3.6.4.1.1. The SR is revised to permit conditions during which the secondary containment may not meet the SR acceptance criterion for a period of up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if an analysis demonstrates that one standby gas treatment (SGT) subsystem remains capable of establishing the required secondary containment vacuum.

EGC has evaluated the proposed change against the criteria of 10 CFR 50.92(c) to determine if the proposed changes result in any significant hazards. The following is the evaluation of each of the 10 CFR 50.92(c) criteria:

1.

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

Response

No The increase in the total MSIV leakage rate limit has been evaluated in a revision to the radiological consequence analysis of the Loss of Coolant Accident (LOCA). Based on the results of the analysis, it has been demonstrated that, with the requested change, the dose consequences of this limiting Design Basis Accident (DBA) are within the acceptance criteria provided by the NRC for use with the Alternative Source Term (AST) methodology in 10 CFR 50.67 and 10 CFR 50, Appendix A, GDC 19. Additional guidance is provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis

ATTACHMENT 1 Evaluation of Proposed Changes Page 25 Accidents at Nuclear Power Reactors," Regulatory Issue Summary (RIS) 2006-04, and Standard Review Plan (SRP) Section 15.0.1.

The proposed change to the MSIV leakage limit does not involve physical change to any plant structure, system, or component. The proposed change does not affect the normal design or operation of the facility before the accident; rather, it affects leakage limit assumptions that constitute inputs to the evaluation of the consequences. The radiological consequences of the analyzed LOCA have been evaluated using the plant licensing basis for this accident. The resulting doses are higher than the previously approved AST doses. However, adequate margin to the regulatory limits specified in 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room operator doses is still available. Thus, the results conclude that the control room and offsite doses remain within applicable regulatory limits.

The effect of the proposed changes on Environmental Qualification (EQ) and vital area access doses have also been evaluated. The proposed increase in MSIV leak rate does not require any new components to be evaluated for inclusion in the EQ program and all components currently in the program remain qualified for their environments. The dose rates and doses to personnel performing vital area tasks post-LOCA remain within acceptance criteria with the proposed change.

In addition, the proposed change to SR 3.6.4.1.1 addresses short-duration conditions during which the secondary containment vacuum requirement is not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four-hour Completion Time (i.e., allowed outage time) for an inoperable secondary containment. In addition, the proposed change provides an alternative means to ensure the secondary containment safety function is met. As a result, the consequences of an accident previously evaluated are not significantly increased.

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

2.

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

Response

No The change in the MSIV leakage rate limits does not affect the design, functional performance, or normal operation of the facility. Similarly, it does not affect the design or operation of any component in the facility such that new equipment failure modes are created. This is supported by operating experience at other EGC sites that have increased their MSIV leakage limits. As such the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

In addition, the proposed change to SR 3.6.4.1.1 does not alter the protection system design, create new failure modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be

ATTACHMENT 1 Evaluation of Proposed Changes Page 26 installed. Consequently, there are no new initiators that could result in a new or different kind of accident.

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

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response

No This proposed license amendment involves changes in the MSIV leakage rate limits. The revised leakage rate limits are used in the reanalysis of the LOCA radiological consequences.

The analysis has been performed using conservative methodologies. Safety margins and analytical conservatisms have been evaluated and found acceptable. The analyzed LOCA event has been carefully selected and margin has been retained to ensure that the analysis adequately bounds postulated event scenario. The dose consequences of this limiting event are within the acceptance criteria presented in 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room operator doses. The margin of safety is that provided by meeting the applicable regulatory limits.

In addition, the proposed change to SR 3.6.4.1.1 addresses short-duration conditions during which the secondary containment vacuum requirement is not met. Conditions in which the secondary containment vacuum is less than the required vacuum are acceptable provided the conditions do not affect the ability of the SGT (standby gas treatment) System to establish the required secondary containment vacuum under post-accident conditions within the time assumed in the accident analysis. This condition is incorporated in the proposed change by requiring an analysis of actual environmental and secondary containment pressure conditions to confirm the capability of the SGT System is maintained within the assumptions of the accident analysis. Therefore, the safety function of the secondary containment is not affected.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

4.3 Conclusion Based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's 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.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed

ATTACHMENT 1 Evaluation of Proposed Changes Page 27 amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1.

Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements" (CAC No. MF5125), dated September 21, 2017 (ADAMS Accession No. ML17236A368) 6.2.

Letter from B. Purnell (U.S. NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Supplemental Information Needed for Acceptance of License Amendment Request to Revise Technical Specification Requirements for Secondary Containment (EPID L-2017-LLA-0379)," dated January 9, 2018 (ADAMS Accession No. ML17353A949) 6.3.

Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Withdrawal of Application to Revise Technical Specifications to Adopt TSTF-551, 'Revise Secondary Containment Surveillance Requirements'," dated January 24, 2018 (ADAMS Accession No. ML18024B022) 6.4.

Letter from P.R. Simpson (Exelon Generation Company) to U.S. Nuclear Regulatory Commission, "Additional Information Supporting the Request for License Amendment Related to Application of Alternative Source Term," dated August 22, 2005 (ADAMS Accession No. ML052430273) 6.5.

Letter from Maitri Banerjee (U.S. NRC) to Christopher M. Crane (Exelon Generation Company), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Re: Adoption of Alternative Source Term Methodology (TAC Nos. MB6530, MB6531, MB6532, MB6533, MC8275, MC8276, MC8277 and MC8278)," dated September 11, 2006 (ADAMS Accession No. ML062070292) 6.6.

Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792) 6.7.

S.L. Humphreys, et al., NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation," (originally published April 1998) (ADAMS Accession No. ML15092A284) 6.8.

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)

ATTACHMENT 1 Evaluation of Proposed Changes Page 28 6.9.

J.E. Cline & Associates, Inc., "MSIV Leakage Iodine Transport Analysis," Letter Report dated March 26, 1991 (ADAMS Accession No. ML003683718) 6.10. Dresden Nuclear Power Station, Units 2 and 3 Technical Specifications, as revised through Amendment 260/253 6.11. Dresden Nuclear Power Station Technical Requirements Manual 3.6.a, "Drywell Spray,"

Revision 0 6.12. Dresden Nuclear Power Station Updated Final Safety Analysis Report (UFSAR),

Revision 13 6.13. Letter from B. Purnell (U.S. NRC) to B. Hanson (Exelon Generation Company, LLC),

"Dresden Nuclear Power Station, Units 2 and 3; Lasalle County Station, Units 1 and 2; and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments to Revise Surveillance Requirement for Secondary Containment Access Doors (CAC Nos.

MF7325-MF7330)," dated February 16, 2017 (ADAMS Accession No. ML17037D212) 6.14. U.S. NRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC Generic Letter 99-02, dated June 3, 1999 (ADAMS Accession No. ML082350935) and corresponding Errata dated August 23, 1999 (ADAMS Accession No. ML031110094) 6.15. NUREG-0800, Standard Review Plan Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," dated March 2007 (ADAMS Accession No. ML070190178) 6.16. NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," dated March 7, 2006 (ADAMS Accession No. ML053460347) 6.17. Fundamentals Handbook, ASHRAE, 1997 6.18. NUREG/CR-0009, "Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels," A.K. Postma, R.R. Sherry, and P.S. Tam, October 1978 (ADAMS Accession No. 7812050096) 6.19. BNWL-1457, "Natural Transport Effects on Fission Product Behavior in the Containment Systems Experiment," R. K. Hilliard and L. F. Coleman, December 1970 available for download as of September 2019 from https://www.osti.gov/biblio/4076997 6.20. WASH-1400 (NUREG-75/014), "An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," Appendices VII, VIII, IX, and X, dated October 1975 (ADAMS Accession No. ML070600376) 6.21. 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 available for download as of September 2019 from https://digital.library.unt.edu/ark:/67531/metadc101034/m2/1/high_res_d/metadc101034.pdf

ATTACHMENT 1 Evaluation of Proposed Changes Page 29 6.22. NUREG-0800, Standard Review Plan Section 15.0, "Introduction - Transient and Accident Analyses," dated March 2007 (ADAMS Accession No. ML070710376) 6.23. Journal of Nuclear Materials, Volume 465, October 2015, Pages 127-134, Fission Product Partitioning in Aerosol Release from Simulated Spent Nuclear Fuel, by F.G. Di Lemma, J.Y. Colle, G. Rasmussen, and R.J.M. Konings.

ATTACHMENT 2 DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3 Docket Nos. 50-237 and 50-249 Facility Operating License Nos. DPR-19 and DPR-25 MARK-UP OF DNPS, UNITS 2 AND 3 TECHNICAL SPECIFICATIONS PAGES

PCIVs 3.6.1.3 Dresden 2 and 3 3.6.1.3-8 Amendment No. 256/249 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.10 Verify the leakage rate through each MSIV leakage path is 34 scfh when tested at 25 psig, and the combined leakage rate for all MSIV leakage paths is 86 scfh when tested at 25 psig.

In accordance with the Primary Containment Leakage Rate Testing Program 62.4 scfh for Unit 2 and 78 scfh for Unit 3 156 scfh for Unit 2 and 218 scfh for Unit 3

Drywell Spray 3.6.2.6 Dresden 2 and 3 3.6.2.6-1 Amendment No.

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 subsystem inoperable.

A.1 Restore drywell spray subsystem to OPERABLE status.

7 days B. Two drywell spray subsystems inoperable.

B.1 Restore one drywell spray subsystem to OPERABLE status.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

Drywell Spray 3.6.2.6 Dresden 2 and 3 3.6.2.6-2 Amendment No.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify 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.

In accordance with the Surveillance Frequency Control 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 gas accumulation are sufficiently filled with water.

In accordance with the Surveillance Frequency Control Program

Secondary Containment 3.6.4.1 Dresden 2 and 3 3.6.4.1-2 Amendment No. 253/246 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 0.25 inch of vacuum water gauge.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify one secondary containment access door in each access opening is closed, except when the access opening is being used for entry and exit.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be maintained 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate 4000 cfm.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment equipment hatches are closed and sealed.

In accordance with the Surveillance Frequency Control Program

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

Not required to be met for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.

ATTACHMENT 3 DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3 Docket Nos. 50-237 and 50-249 Facility Operating License Nos. DPR-19 and DPR-25 CLEAN DNPS, UNITS 2 AND 3 TECHNICAL SPECIFICATIONS PAGES

PCIVs 3.6.1.3 Dresden 2 and 3 3.6.1.3-8 Amendment No. 256/249 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.

In accordance with the Surveillance Frequency Control Program SR 3.6.1.3.10 Verify the leakage rate through each MSIV leakage path is 62.4 scfh for Unit 2 and 78 scfh for Unit 3 when tested at 25 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.

In accordance with the Primary Containment Leakage Rate Testing Program

Drywell Spray 3.6.2.6 Dresden 2 and 3 3.6.2.6-1 Amendment No.

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 subsystem inoperable.

A.1 Restore drywell spray subsystem to OPERABLE status.

7 days B. Two drywell spray subsystems inoperable.

B.1 Restore one drywell spray subsystem to OPERABLE status.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

Drywell Spray 3.6.2.6 Dresden 2 and 3 3.6.2.6-2 Amendment No.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify 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.

In accordance with the Surveillance Frequency Control 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 gas accumulation are sufficiently filled with water.

In accordance with the Surveillance Frequency Control Program

Secondary Containment 3.6.4.1 Dresden 2 and 3 3.6.4.1-2 Amendment No. 253/246 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1

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

Not required to be met for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.

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

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify one secondary containment access door in each access opening is closed, except when the access opening is being used for entry and exit.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be maintained 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate 4000 cfm.

In accordance with the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment equipment hatches are closed and sealed.

In accordance with the Surveillance Frequency Control Program

ATTACHMENT 4 DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3 Docket Nos. 50-237 and 50-249 Facility Operating License Nos. DPR-19 and DPR-25 MARK-UP OF DNPS, UNITS 2 AND 3 TECHNICAL SPECIFICATIONS BASES PAGES (For Information Only)

Primary Containment B 3.6.1.1 Dresden 2 and 3 B 3.6.1.1-2 Revision 31 BASES BACKGROUND This Specification ensures that the performance of the (continued) primary containment, in the event of a Design Basis Accident (DBA), meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J, Option B (Ref. 3), as modified by approved exemptions.

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

The maximum allowable leakage rate for the primary containment (La) is 3.0% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at containment pressure of 43.9 psig.

Primary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Primary containment OPERABILITY is maintained by limiting leakage to 1.0 La, except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met. In addition, the leakage from the drywell to the suppression chamber must be limited to (continued) and is reduced to 1.5% by weight at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a LOCA. MSIV leakage is not considered part of La during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a LOCA

PCIVs B 3.6.1.3 Dresden 2 and 3 B 3.6.1.3-14 Revision 69 BASES SURVEILLANCE SR 3.6.1.3.9 REQUIREMENTS (continued)

The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Other administrative controls, such as those that limit the shelf life and operating life, as applicable, of the explosive charges must be followed.

SR 3.6.1.3.10 The analyses in References 2 and 3 are based on leakage that is less than the specified leakage rate. In accordance with the Primary Containment Leakage Rate Testing Program, the as-left leakage rate of each main steam isolation valve path is assumed to be the maximum pathway leakage (larger leakage of two valves in series), and the as-found leakage rate of each main steam isolation valve path is assumed to be the minimum pathway leakage (smaller of either the inboard or outboard isolation valves individual leakage rates). The combined leakage rate limit for all MSIV leakage paths must be < 86 scfh when tested at > 25 psig for both as-left and as-found leakage rate tests. Additionally, the leakage rate limit through each MSIV leakage path is < 34 scfh when tested at > 25 psig. These values correspond to a combined leakage rate of 150 scfh and an individual MSIV leakage rate of 60 scfh, when tested at 43.9 psig. This ensures that MSIV leakage is properly accounted for in determining the overall impacts of primary containment leakage. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

MSIV leakage is considered part of La.

(continued) 62.4 scfh for Unit 2 and 78 scfh for Unit 3 156 scfh for Unit 2 and 218 scfh for Unit 3 250 scfh for Unit 2 and 350 scfh for Unit 3 100 scfh for Unit 2 and 125 scfh for Unit 3 and unit-specific atmospheric dispersion factors.

Drywell Spray B 3.6.2.6 Dresden 2 and 3 B 3.6.1.3-1 Revision 0 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.6 Drywell Spray BASES BACKGROUND The drywell spray subsystem of the low pressure coolant injection (LPCI) system is operated post-loss-of-coolant accident (LOCA) to remove inorganic iodines and particulates from the drywell atmosphere by washing, or scrubbing, them into the suppression pool.

Each of the two drywell spray subsystems contain two pumps, one heat exchanger, drywell spray valves, and a spray header inside the drywell. Each drywell spray subsystem is capable of recirculating water from the suppression pool through a heat exchanger and dispersed through the drywell spray nozzles. The spray then effects a scrubbing or washing of the drywell atmosphere.

The LOCA radiological dose analysis credits the drywell spray subsystem for scrubbing radionuclides from the drywell air space.

The containment cooling mode of LPCI, which includes drywell spray, is described in the UFSAR, Reference 1.

APPLICABLE The drywell spray is credited post-LOCA for scrubbing SAFETY ANALYSES inorganic iodines and particulates from the drywell atmosphere. This function reduces the amount of airborne activity available for leakage from the drywell to ensure that the radiological consequences from the accident remain within the limits of 10 CFR 50.67 (Ref. 4). The drywell spray can also be used to reduce the temperature and pressure in the drywell, which reduces the leak rate of airborne activity from primary containment. However, drywell spray is not required to maintain the drywell temperatures and pressures below the design limits.

Reference 2 contains the results of the analysis used to predict the effects of drywell spray on the post-accident primary containment atmosphere.

The drywell spray subsystem satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

Drywell Spray B 3.6.2.6 Dresden 2 and 3 B 3.6.1.3-2 Revision 0 BASES (continued)

LCO In the event of a Design Basis Accident (DBA), a minimum of one drywell spray subsystem using one LPCI pump is required to adequately scrub the inorganic iodines and particulates from the primary containment atmosphere. To ensure that these requirements are met, two drywell spray subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. A drywell spray subsystem is OPERABLE when one of the pumps and associated piping, valves, instrumentation, and controls are OPERABLE.

Management of gas voids is important to drywell spray subsystem OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, a DBA could release fission products into the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.

Therefore, maintaining drywell spray subsystems OPERABLE is not required in MODE 4 or 5.

ACTIONS A.1 With one drywell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.

In this condition, the remaining OPERABLE drywell spray subsystem is adequate to perform the primary containment fission product scrubbing function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in the loss of the scrubbing capability of the drywell spray subsystem. The 7-day Completion Time was chosen in light of the redundant drywell spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.

B.1 With both drywell spray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this condition, there is a substantial loss of the fission product scrubbing function of the drywell spray (continued)

Drywell Spray B 3.6.2.6 Dresden 2 and 3 B 3.6.1.3-3 Revision 0 BASES ACTIONS B.1 (continued) system. The 8-hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA.

C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.2.6.1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the drywell spray mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the non-accident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the drywell spray mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

Drywell Spray B 3.6.2.6 Dresden 2 and 3 B 3.6.1.3-4 Revision 0 BASES SURVEILLANCE SR 3.6.2.6.2 REQUIREMENTS (continued)

This surveillance is performed to verify that the spray nozzles are not obstructed and that spray flow will be provided when required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.2.6.3 Drywell spray subsystem associated piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the drywell spray subsystems and may also prevent water hammer and pump cavitation.

Selection of drywell spray subsystem locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The drywell spray subsystem is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the drywell spray subsystem is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met.

Accumulated gas should be eliminated or brought within the acceptance criteria limits.

(continued)

Drywell Spray B 3.6.2.6 Dresden 2 and 3 B 3.6.1.3-5 Revision 0 BASES SURVEILLANCE SR 3.6.2.6.3 (continued)

REQUIREMENTS Drywell spray subsystem locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative subset of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES

1.

UFSAR, Section 6.2.2.2.

2.

UFSAR, Section 15.6.5.

3.

10 CFR 50.36(c)(2)(ii).

4.

10 CFR 50.67, "Accident Source Term."

Secondary Containment B 3.6.4.1 Dresden 2 and 3 B 3.6.4.1-4 Revision 75 BASES ACTIONS C.1 (continued)

Movement of recently irradiated fuel assemblies in the secondary containment can be postulated to cause significant fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of this activity shall not preclude completing an action that involves moving a component to a safe position.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

The SR is modified by a Note which states the SR is not required to be met for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if an analysis demonstrates that one SGT subsystem remains capable of establishing the required secondary containment vacuum. Use of the Note is expected to be infrequent but may be necessitated by situations in which secondary containment vacuum may be less than the required containment vacuum, such as, but not limited to, wind gusts or failure or change of operating normal ventilation subsystems. These conditions do not indicate any change in the leak tightness of the secondary containment boundary. The analysis (Ref. 3) should consider the actual conditions (equipment configuration, temperature, atmospheric pressure, wind conditions, measured secondary containment vacuum, etc.) to determine whether, if an accident requiring secondary containment to be OPERABLE were to occur, one train of SGT could establish the assumed secondary containment vacuum within the time assumed in the accident analysis. If so, the SR may be considered met for a period up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit is based on the expected short duration of the situations when the Note would be applied.

Secondary Containment B 3.6.4.1 Dresden 2 and 3 B 3.6.4.1-6 Revision 66 BASES SURVEILLANCE SR 3.6.4.1.3 (continued)

REQUIREMENTS can be maintained 0.25 inches of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate 4000 cfm.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months test period allows secondary containment to be in thermal equilibrium at steady state conditions. The primary purpose of the SR is to ensure secondary containment boundary integrity. The secondary purpose of the SR is to ensure that the SGT subsystem being tested functions as designed. There is a separate LCO with Surveillance Requirements that serves the primary purpose of ensuring OPERABILITY of the SGT System. This SR need not be performed with each SGT subsystem. The SGT subsystem used for this Surveillance is staggered to ensure that in addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform this test. The inoperability of the SGT System does not necessarily constitute a failure of this Surveillance relative to secondary containment OPERABILITY.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.4.1.4 Verifying that secondary containment equipment hatches are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur and provides adequate assurance that exfiltration from the secondary containment will not occur.

In this application, the term "sealed" has no connotation of leak tightness. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1.

UFSAR, Section 15.6.5.

2.

NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

3. DRE19-0015, Revision 0a, Dresden Units 2

& 3 Secondary Containment Drawdown Analysis, October 2019.

ATTACHMENT 5 Proposed Exemption to Certain 10 CFR 50, Appendix J, Option B Requirements Page 1

1.

SPECIFIC EXEMPTION REQUEST 10 CFR 50.54(o) requires that primary reactor containments be subjected to the requirements of 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J specifies the leakage rate test requirements, schedules, and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components that penetrate containment.

10 CFR 50, Appendix J, Option B, "Performance-Based Requirements," Paragraphs III.A, "Type A Test," requires that the overall integrated leakage rate (i.e., Type A test) must not exceed the allowable leakage (La) with margin, as specified in the Technical Specifications (TS).

Paragraph III.B, "Type B and C Tests," requires the sum of the leakage of Type B and Type C local leakage rate tests to be less than the performance criterion (La) with margin as specified in the TS. As specified in the 10 CFR 50 Appendix J definitions, the overall integrated leakage rate is the total leakage rate from all tested leakage paths, including contributions from containment welds, valves, fittings, and components that penetrate the containment. The four main steam line penetrations are included in this definition of overall integrated leakage rate.

The Main Steam Isolation Valve (MSIV) leakage pathway consists of the combined leakage of the four main steam lines where each line contains two MSIVs in series and is tested in accordance with TS Surveillance Requirement (SR) 3.6.1.3.10.

In accordance with 10 CFR 50.12, "Specific exemptions," Exelon Generation Company, LLC (EGC) requests an exemption from the requirements of 10 CFR 50, Appendix J, Option B, Paragraphs III.A and III.B for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The proposed exemption would permit exclusion of the MSIV leakage contribution from the overall integrated leakage rate Type A test measurement and from the sum of the leakage rates from Type B and Type C tests. This exemption will allow the leakage testing to be performed in a manner consistent with the way MSIV leakage is modeled in the revised radiological consequence analysis provided in Enclosure B.

2.

BASIS FOR EXEMPTION REQUEST 10 CFR 50, Appendix J testing ensures primary containment leakage following a design basis loss-of-coolant accident (LOCA) will be within the allowable leakage limits specified in plant TS and assumed in the safety analysis for determining radiological consequences. The DNPS, Units 2 and 3 Primary Containment Leakage Rate Testing Program as described in Technical Specification 5.5.12, "Primary Containment Leakage Rate Testing Program," complies with 10 CFR 50, Appendix J, Option B provisions for Type A, Type B and Type C tests. MSIV leakage currently is included in the Type A overall containment integrated leakage rate total and added to the combined Type B and C leakage rate total. However, the radiological consequences of MSIV leakage are modeled as a separate primary containment release path to the environment that bypasses secondary containment and therefore it is not filtered through the standby gas treatment system like other containment leakage. The LOCA dose calculation assumes all MSIV leakage migrates to the Turbine Building and then to the environment.

EGC requests a permanent exemption for DNPS, Units 2 and 3 from: 1) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A, to allow exclusion of the MSIV leakage from the overall integrated leakage rate measured when performing a Type A test, and 2) the

ATTACHMENT 5 Proposed Exemption to Certain 10 CFR 50, Appendix J, Option B Requirements Page 2 requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.B, to allow exclusion of the MSIV leakage from the combined leakage rate of all penetrations and valves subject to Type B and C tests. Approval of this exemption request will more closely align the DNPS, Units 2 and 3 Primary Containment Leakage Rate Testing Program with the assumptions used in the accident consequence analyses.

3.

EXEMPTION JUSTIFICATION

a.

The exemption is authorized by law 10 CFR 50.12(a)(1) requires a demonstration that an exemption from NRC regulations is authorized by law. The proposed exemption is authorized by law and has been previously granted to other licensees as described in Section 5 below.

b.

The exemption will not present undue risk to public health and safety 10 CFR 50.12(a)(1) requires a demonstration that the granting of an exemption from the requirement in question "will not present an undue risk to the public health and safety." The exemption presents no undue risk to public health and safety. MSIV leakage for the revised DNPS Design Basis Accident (DBA) analysis has been accounted for separately from the overall leakage associated with the primary containment boundary (i.e., Type A) and local leakage rate total (i.e., Type B and C) (e.g., MSIV leakage is separate from the maximum TS allowable leakage, La). As such, the inclusion of MSIV leakage as part of Type A and as part of Type B and C test results is not necessary to ensure the actual radiological consequences of DBAs remain below the regulatory limit. Since the exemption will not result in a significant change to the previously evaluated consequences associated with DBAs, the proposed exemption presents no undue risk to public health and safety.

c.

The exemption is consistent with the common defense and security The term "common defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested exemption to remove MSIV leakage from La will not affect any of these and, thus, such grants are consistent with the common defense and security. Further, the potential impact on public health and safety has been determined to be inconsequential.

4.

SPECIAL CIRCUMSTANCES 10 CFR 50.12 states that the NRC will not consider granting an exemption unless special circumstances are present. This request meets the criterion of a special circumstance as defined in 10 CFR 50.12(a)(2)(ii), which states, "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule."

The pertinent rule is 10 CFR 50, Appendix J, Option B, Paragraphs III.A and III.B, which state in part that:

ATTACHMENT 5 Proposed Exemption to Certain 10 CFR 50, Appendix J, Option B Requirements Page 3 a) Type A Test leakage rate must not exceed the allowable leakage rate (La) with margin, as specified in the TS.

b) Type B and C Tests must demonstrate that the sum of the leakage rates at accident pressure of Type B tests, and pathway leakage rates from Type C tests, is less than the performance criterion (La) with margin, as specified in the TS.

The underlying purpose of the rule is to ensure that the radiological consequences of DBAs remain below the applicable regulatory limits and are supported by the actual, periodic measurement of containment leakage (Type A) and local leakage rate measurement (Type B and C). Although Type A and Type B and C leakage tests are defined as a measurement of those leakages, inclusion of the MSIV leakage results in double counting, once as a part of the actual containment leakage and again as part of MSIV leakage used in dose calculations. This exemption resolves the special circumstance in which requiring inclusion of MSIV leakage in the Type A and Type B and C leakage is not necessary to achieve the underlying purpose of the rule.

5.

PRECEDENTS The requested exemption will align DNPS with the common industry practice of monitoring MSIV leakage separate from the station allowable leakage, La, totals. Examples of recent similar exemptions requested granted by the NRC include:

"Monticello Nuclear Generating Plant - Issuance of Exemption to Certain 10 CFR Part 50, Appendix J, Requirements," dated December 7, 2006 (Reference 1)

"Vermont Yankee Nuclear Power Station - Issuance of Exemption from 10 CFR Part 50, Appendix J," dated March 17, 2005 (Reference 2)

"Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Exemption from 10 CFR Part 50, Appendix J," dated March 14, 2000 (Reference 3)

6.

ENVIRONMENTAL IMPACT In accordance with 10 CFR 51.30 and 10 CFR 51.32, the following information is provided in support of an environmental assessment and finding of no significant impact for the proposed action.

The exemption does not cause additional construction or operational activities to be conducted that may significantly affect the environment. No plant configuration changes are required. The exemption does not result in an increase in adverse environmental impact previously evaluated, does not result in a significant change to normal offsite effluents or power levels, and does not affect any matter not previously reviewed by the NRC which may have a significant adverse environmental impact.

The exemption does not alter the land use for the plant; water uses or impacts on water quality; air or ambient air quality. It does not affect the ecology of the site and vicinity and does not

ATTACHMENT 5 Proposed Exemption to Certain 10 CFR 50, Appendix J, Option B Requirements Page 4 affect the noise emitted by the station. In addition, the proposed exemption request does not affect non-radiological plant effluents and has no other environmental impact.

Therefore, the proposed exemption does not affect the analysis of environmental impacts described in the environmental report.

7.

CONCLUSION As demonstrated above, this request for an exemption from the requirements of 10 CFR 54(o) and the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A, and Paragraph III.B meet the criteria of 10 CFR 50.12 for specific exemptions. Specifically, the requested exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. In addition, the special circumstances described in 10 CFR 50.12(a)(2)(ii) are present and warrant issuance of the exemption.

8.

REFERENCES

1. Letter from P.S. Tam (U.S. NRC) to J.T. Conway (Nuclear Management Company. LLC),

"Monticello Nuclear Generating Plant - Issuance of Exemption to Certain 10 CFR Part 50, Appendix J, Requirements," dated December 7, 2006 (ADAMS Accession No. ML062410507)

2. Letter from R.B. Ennis (U.S. NRC) to M. Kansler (Entergy Nuclear Operations Inc),

"Vermont Yankee Nuclear Power Station - Issuance of Exemption from 10 CFR Part 50, Appendix J," dated March 17, 2005 (ADAMS Accession No. ML041310359)

3. Letter from W.O. Long (U.S. NRC) to J.A. Scalice (Tennessee Valley Authority), "Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Exemption from 10 CFR Part 50, Appendix J," dated March 14, 2000 (ADAMS Accession No. ML003691985)

v t e Letter Sequence Request
EPID:L-2019-LLA-0232, Response to Request for Additional Information for the License Amendment Request to Change Technical Specifications to Increase Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions 27 March 2020 18 August 2020 Answers 6 May 2020 Results Approval, Approval, Approval Other: ML19324F481
MONTHYEAR2020-10-05 [Table View]

ML19294A304

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Subject:

References:

Enclosures:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

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Response

Response

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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