RS-19-023, Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 3.6.4.1.1

Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 3.6.4.1.1

ML19064B369
Person / Time
Site:	Quad Cities
Issue date:	03/05/2019
From:	Simpson P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19064B368	List: ML19064B370 ML19064B371 RS-19-023, Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 3.6.4.1.1
References
RS-19-023
Download: ML19064B369 (51)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.90 RS-19-023 March 5, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 3.6.4.1.1

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

March 5, 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-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 Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2.

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 1 and 218 scfh for Unit 2, respectively. The proposed amendment also revises the leakage rate through each MSIV leakage path from 34 to 62.4 scfh for Unit 1 and 78 scfh for Unit 2, 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 QCNPS alternative source term (AST) methodology previously approved by the NRC in Reference 1. MSIV leakage will also no longer be counted as part of L a aligning QCNPS with the common industry practice of monitoring MSIV leakage separate from the station L a totals. A new TS 3.6.2.6, "Residual Heat Removal (RHR) Drywell Spray" is added to reflect the crediting of drywell spray for fission product removal as part of this 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 QCNPS, which lead to withdrawal of a prior submittal of this change (Reference 4), have now been addressed. 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.

March 5, 2019 U.S. Nuclear Regulatory Commission Page 3 The proposed change has been reviewed by the QCNPS Plant Operations Review Committee, in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed license amendment by March 5, 2020. Once approved, the amendment shall be implemented within 60 days.

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 5th day of March 2019.

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

1. Evaluation of Proposed Changes

2. Mark-up of QCNPS, Units 1 and 2 Technical Specifications Pages

3. Clean QCNPS, Units 1 and 2 Technical Specifications Pages

4. Mark-up of QCNPS, Units 1 and 2 Technical Specifications Bases Pages - For Information Only

Enclosures:

A. QDC-7500-M-2341, Revision 0, Quad Cities Units 1 & 2 Secondary Containment Drawdown Analysis

8. QDC-OOOO-N-1481, Revision 3, Quad Cities Units 1 & 2 Post-LOCA EAB, LPZ, and CR Dose - AST Analysis cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Quad Cities Nuclear Power Station NRC Project Manager, Quad Cities 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 3.6.4.1.1 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Reason for the Proposed Changes 2.2 Description of the Proposed Changes

3.0 TECHNICAL EVALUATION

3.1 Evaluation of Proposed Change to the MSIV Leakage Rate Limits 3.2 New Technical Specifications 3.6.2.6 Residual Heat Removal (RHR) 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

6.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2.

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 1 and 218 scfh for Unit 2, respectively. The proposed amendment also revises the leakage rate through each MSIV leakage path from 34 to 62.4 scfh for Unit 1 and 78 scfh for Unit 2, 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 QCNPS alternative source term (AST) methodology previously approved by the Nuclear Regulatory Commission (NRC) in Reference 6.2. MSIV leakage will also no longer be counted as part of L a aligning QCNPS with the common industry practice of monitoring MSIV leakage separate from the station L a totals.

TS 3.6.2.6, "Residual Heat Removal (RHR) Drywell Spray," is added to reflect the crediting of drywell spray for fission product removal as part of this 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 QCNPS in Reference 6.3, which lead to withdrawal of a prior submittal of this change (Reference 6.4), have now been addressed. The portion of TSTF-551 that modifies SR 3.6.4.1.3 is already incorporated into the QCNPS, Units 1 and 2 TS SR 3.6.4.1.2 and is therefore not included in this license amendment request.

2.0 DETAILED DESCRIPTION The NRC approved the use of AST for the evaluation of the onsite and offsite dose consequences for the following Design Basis Accidents: Loss of Coolant Accident (LOCA),

Control Rod Drop Accident (CRDA), Fuel Handling Accident (FHA), and Main Steam Line Break (MSLB) at QCNPS in Reference 6.2. 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 QCNPS, Units 1 and 2 TS and adopt the portion of TSTF-551 that requires a drawdown time.

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ATTACHMENT 1 Evaluation of Proposed Changes 2.1 Reason for the Proposed Changes 2.1.1 Revising MSIV Leakage Rate Limits The four main steam lines, which penetrate the drywell, are automatically isolated by the MSIVs.

There are two MSIVs on each steam line, one inside containment (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.

Unplanned MSIV repairs are a significant contributor to increased outage duration and unplanned radiation exposure during refueling outages. Increasing the MSIV leakage rate limits in TS 3.6.1.3 will relax operational constraints during outage activities and will improve the overall performance integrity 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. However, the limiting QCNPS design basis accident analysis considering the revised MSIV leakage rate must continue to meet the acceptance criteria of 10 CFR 50.67 for Accident Source Term and 10 CFR 50, Appendix A, GDC 19 for control room (CR) dose consequences.

The change to exclude MSIV leakage from the L a totals will align QCNPS with standard industry practice and allow for a more direct comparison when evaluating L a performance.

2.1.2 New TS 3.6.2.6 Residual Heat Removal (RHR) Drywell Spray The current licensing basis (CLB) does not credit drywell spray for mitigation of any 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 surveillance requirements are being moved from the Technical Requirements Manual (TRM) to the TS.

2.1.3 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 Design Basis Accident (DBA) to ensure the control room operator and offsite doses are within the regulatory and NRC-approved limits. The secondary containment requires support systems to maintain the control volume 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 0.10 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at a flow rate 4000 cfm.

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ATTACHMENT 1 Evaluation of Proposed Changes Secondary containment is a single train system that performs a safety function. Per the discussion 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, "License 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, ACTION (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.

2.2 Description of the Proposed Changes The allowable leakage rate specified in SR 3.6.1.3.10 will be changed to provide unit specific leakage rate limits. As described in Table 3-1, the MSIV leakage rate assumed in the LOCA dose consequence analysis is 250 scfh and 350 scfh, for Unit 1 and 2, respectively, at 43.9 psig.

As described in QCTP 0130-01 page 15 (Reference 6.18), 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 6.19) 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 1, 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 2 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 QCNPS design basis accident. 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 Verify the leakage rate through each MSIV leakage path is 34 scfh when tested at leakage path is 34 62.4 scfh for Unit 1 and 25 psig, and the combined leakage rate for 78 scfh for Unit 2 when tested at 25 psig, and all MSIV leakage paths is 86 scfh when the combined leakage rate for all MSIV tested at 25 psig. leakage paths is 86 156 scfh for Unit 1 and 218 scfh for Unit 2 when tested at 25 psig.

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ATTACHMENT 1 Evaluation of Proposed Changes The TS for drywell spray currently resides in the TRM because the CLB does not credit the system for accident mitigation. TS 3.6.2.6, "Residual Heat Removal (RHR) Drywell Spray" is added as a new TS because the revised LOCA analysis credits drywell spray for accident mitigation. The TS, Limiting Condition of Operation (LCO), applicability, action statements, and SRs are patterned after existing TS 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray." The proposed LCO requires two RHR drywell spray subsystems to be operable.

3.6.2.6 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.6 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

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

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

C. Required Action and -------------------NOTE-----------------

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Page 4

ATTACHMENT 1 Evaluation of Proposed Changes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.6.1 Verify each RHR drywell spray subsystem manual and In accordance with power operated valve in the flow path that is not locked, the Surveillance sealed, or otherwise secured in position, is in the correct Frequency Control position or can be aligned to the correct position. Program SR 3.6.2.6.2 Verify each drywell spray nozzle is unobstructed. In accordance with the Surveillance Frequency Control Program In accordance with SR 3.6.2.6.3 Verify RHR drywell spray subsystem locations the Surveillance susceptible to gas accumulation are sufficiently filled with Frequency Control water. Program 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 0.10 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 0.10-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 --------------------------NOTE--------------------------

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

Verify secondary containment vacuum is 0.10 inch of vacuum water gauge.

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 L a . contains a marked-up version of the QCNPS, Units 1 and 2 TS showing the proposed changes. Attachment 3 provides the revised (clean) TS pages.

Page 5

ATTACHMENT 1 Evaluation of Proposed Changes 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 for RHR Drywell Spray and for Surveillance Frequency Control Program 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 Evaluation of Proposed Change to the MSIV Leakage Rate Limits On September 11, 2006 (Reference 6.2) the NRC issued Amendments Nos. 233 and 229 to the Renewed Facility Operating Licenses for QCNPS, Units 1 and 2, respectively. These amendments adopt the full implementation of the AST methodology in accordance with 10 CFR 50.67, "Accident Source Term." At that time QCNPS 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 6.6) radiological consequence analysis of the design basis LOCA.

The revised analysis was performed in accordance with Regulatory Guide (RG) 1.183 (Reference 6.5) 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. 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. A discussion of the reason for each change is provided below Table 3-1.

Table 3-1: Summary of LOCA Analysis Revisions Design Input Parameter Current Licensing Revised Value Additional Value Information Reduction of MSIV and Not Credited Credited Enclosure B, Containment Leak Rate by Section 2.1.3 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA with Respect to RADTRAD Flow Rates Aerosol Deposition in Not Credited for MSL Credited for MSL Enclosure B, Horizontal MSLs Upstream of containing MSIV that containing MSIV that Section 7.3 Inboard MSIV failed to close failed to close Fraction of Containment 0.0 100% (during Enclosure B, Leakage that Bypasses the drawdown period) Sections 2.3 Standby Gas Treatment (SGT) 0% (following & 5.3.2.8 System due to High Winds drawdown period)

Percentage of Engineered 100% 0% (during Enclosure B, Safety Feature (ESF) and drawdown period) Section Containment Leakage that is 100% (following 5.4.10 filtered by the SGT System drawdown period)

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ATTACHMENT 1 Evaluation of Proposed Changes Table 3-1: Summary of LOCA Analysis Revisions Design Input Parameter Current Licensing Revised Value Additional Value Information SGT System Exhaust Elemental Iodine: Elemental Iodine: Enclosure B, Charcoal Filter Efficiencies 80% 90% Section 7.9 Organic Iodide: Organic Iodide:

80% 90%

MSIV Leak Rate Through All 150 scfh @ 43.9 psig 250 scfh @ 43.9 psig Enclosure B, Four Lines for 0 to 30 days for 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Sections Unit 1 and 350 scfh 2.3.2, 4.6.6,

@ 43.9 psig for 0 to & 5.5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Unit 2 MSIV Leakage per Line (scfh) 60 60 30 0 (See Figure 2) Unit 1: 100 100 50 0 (See Figure 1)

Unit 2: 125 125 100 0

Elemental Iodine Removal 50% Removal Time Dependent Enclosure B, Efficiency in MSLs Efficiency Credited Removal Efficiency Section 2.1.3 for Accident Duration Control Room (CR) Unfiltered 60,000 cfm (includes 4000 cfm (includes Enclosure B, Inleakage during Normal ingress/egress ingress/egress Section 4.8.6 Operation inleakage of 10 cfm) inleakage of 10 cfm)

Particulate (Aerosol) Powers 10 Not credited Enclosure B, Deposition/Plateout Model in percentile Model Section 6-1 Containment Elemental Iodine Removal in Credited Not credited Enclosure B, Containment via Natural Section 2.1.3 Deposition & Table 6-1 Total Containment Leakage 3% volume/day 3% volume/day Enclosure B, minus MSIV leakage Section 5.8 Drywell Spray Not credited Credited Enclosure B, Sections 2.1.3, 4.6.7,

& 7.11 Reduction of MSIV and Containment Leak Rate by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA. Reduction in the containment leakage and MSIV leakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% 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 Page 7

ATTACHMENT 1 Evaluation of Proposed Changes 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the event is conservative with respect to the actual expected leakage rate. As explained above, this reduction on affects the flow rated modeled in RADTRAD. The aerosol removal efficiencies are conservatively calculated based on the non-reduced flow rates.

Aerosol Deposition in Horizontal Main Steam Lines Upstream of Inboard MSIV The QCNPS 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.

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 inboard MSLs to withstand the SSE, the horizontal pipe surface area and volume is credited in the revised aerosol removal analysis.

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 an assumed main steam line pipe break just upstream of a MSIV. The initiating event is a large pipe break of a recirculation suction line 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, so the LOCA dose analysis similarly does not need to consider multiple simultaneous pipe breaks.

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. This analysis assumes that following an accident, the secondary containment ventilation system isolates, and the SGT System is initiated. The results of this analysis determined that the secondary containment drawdown is achieved in 23 minutes. The 23-minute secondary containment drawdown time is based on operation of a single SGT System train (assuming failure of an electrical division). For conservatism, the Page 8

ATTACHMENT 1 Evaluation of Proposed Changes 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 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 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 6.7, 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 6.8 requires a safety factor of at least 2 be used to determine the filter efficiencies credited in the design basis 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 1 and 350 scfh for Unit 2. Unit 2 is analyzed for a higher leakage rate than Unit 1 because the Unit 2 MSIV to control room ground release atmospheric dispersion coefficients are lower than the Unit 1 values. All other inputs are common to both Units 1 and 2. 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 Page 9

ATTACHMENT 1 Evaluation of Proposed Changes 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 1 and 350 scfh for Unit 2 is assumed to occur as follows:

1) 100 scfh for Unit 1 and 125 scfh for Unit 2 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 1 and 125 scfh for Unit 2. The deposition removal of aerosol in the horizontal pipe, and the deposition removal of elemental iodine in both the horizontal and vertical pipes, are credited in the steam line between the RPV nozzle and outboard MSIV.

3) 50 scfh through second intact steam line for Unit 1 and 100 scfh for Unit 2. 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 1 model during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.

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ENVIRONMENT DRYWELL Path 1 -

Compartment 2 Path 2 - MSL 1 With MSIV Drywell to Failed MSL 1 RPV Failure to environment MSL 1 With MSIV Failure Nozzle to outboard (100 scfh) - includes RPV nozzle Turbine (100 scfh) MSIV to outboard MSIV removal Building No Mixing Path 3 - Compartment 3 Path 4 - Drywell to Path 5 - Intact Compartment 4 Drywell to Intact MSL 1 Intact MSL 1 MSL 1 to Intact MSL 1 RPV Intact MSL 1 RPV Nozzle to (100 scfh) environment inboard MSIV to (100 scfh) inboard MSIV - includes RPV (100 scfh) outboard MSIV nozzle to inboard - includes inboard MSIV removal MSIV to outboard MSIV removal Path 6 - Compartment 5 Path 7 - Drywell Compartment 6 Path 8 - Intact Drywell to Intact MSL 2 to Intact MSL 2 Intact MSL 2 MSL 2 to Intact RPV Nozzle to (50 scfh) inboard MSIV to environment MSL 2 inboard MSIV - includes RPV outboard MSIV (50 scfh)

(50 scfh) nozzle to inboard - includes inboard MSIV removal MSIV to outboard MSIV removal Figure 1: Revised MSIV Leakage RADTRAD Nodalization Page 11

ENVIRONMENT DRYWELL Path 1 - Drywell Compartment 2 Path 2 - MSL 1 With MSIV Failure to to MSL 1 With Failed MSL 1 RPV environment (60 scfh) - includes inboard MSIV Failure Turbine Nozzle to outboard MSIV MSIV to outboard MSIV removal (60 scfh) Building Path 3 - Compartment 3 Path 4 - Drywell to Compartment 4 Path 5 - Intact MSL No Drywell to Intact MSL 1 Intact MSL 1 (60 scfh) Intact MSL 1 1 to environment Mixing Intact MSL 1 RPV Nozzle to - includes RPV nozzle RPV inboard (60 scfh)

(60 scfh) inboard MSIV to inboard MSIV MSIV to - includes inboard removal outboard MSIV MSIV to outboard MSIV removal Path 6 - Compartment 5 Path 7 - Drywell to Compartment 6 Path 8 - Intact Drywell to Intact MSL 2 Intact MSL 2 (30 scfh) Intact MSL 2 MSL 2 to Intact MSL 2 RPV Nozzle to - includes RPV inboard MSIV to environment (30 scfh) inboard MSIV nozzle to inboard outboard MSIV (30 scfh) - includes MSIV removal inboard MSIV to outboard MSIV removal Figure 2: Current Licensing Basis MSIV Leakage RADTRAD Nodalization Page 12

ATTACHMENT 1 Evaluation of Proposed Changes 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 6.5) indicates that the methodology given in Reference 6.11 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The time steps used are 2, 8, 24, 48, 72, 96, 240, and 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.

Sensitivity runs show that modeling time dependent elemental iodine removal is conservative because it leads to less removal than the 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 824 cfm (778 +/- 46 cfm) during isolation - recirculation mode. The tracer gas test was last performed in February 2018, 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 824 cfm is bounded by the modeled 4,000 cfm.

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 6.14). 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 32,430 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.

Page 13

ATTACHMENT 1 Evaluation of Proposed Changes 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.

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 6.9). The TRM indicates that once every 10 years, the spray headers and nozzles are air tested in the drywell. This test verifies that a flow path exists through the spray header and nozzles and thereby verifies its operational status. QCNPS TRM 3.6.a requires (Reference 6.10), "Two RHR 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.

QCNPS UFSAR Section 6.2.2.2 states that the containment cooling mode of Residual Heat Removal (RHR) is a safety function and consists of two cooling functions: containment spray which consists of drywell spray and suppression chamber spray and suppression pool cooling.

All containment cooling functions are manually initiated. All equipment and piping in the RHR system that feeds the containment spray nozzles are safety-related. Based on the above discussion, even though using the drywell spray system for scrubbing radionuclides from the drywell air space was not considered a safety-related function as part of the original design basis, the system is currently credited in the accident analysis for removing drywell heat to lower drywell temperature and pressures after the peak containment pressure occurs.

Drywell sprays are assumed to start 10 minutes following event initiation and continue for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operators are directed to start drywell sprays when torus pressure is above 5 psig and the drywell pressure and temperatures are above the drywell spray initiation limit curve. Both criteria are met before 10 minutes following a LOCA. Reasonable assurance of the timeliness Page 14

ATTACHMENT 1 Evaluation of Proposed Changes of this manual action is provided by the Operator Response Time Program. This program directs the operator to manually initiate RHR containment cooling mode (containment spray) for a design basis LOCA at 10 minutes after the event. 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.

The drywell low pressure / containment spray inhibited instruments alarm at 1 psig. Per the containment analysis, the drywell remains above 7 psig until at least 40,000 seconds (11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />) following the event, so even though the sprays can continue as long as containment pressure is above 0 psig, containment sprays are assumed to stop at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to conservatively bound the anticipated operator action to continue sprays until containment 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 <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and aerosol iodine mass reaches a DF of 50 at 2.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

After 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> the elemental iodine removal via spray is terminated and after 2.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the aerosol removal coefficient is reduced to 1.5 hr-1 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> post-accident when the drywell spray is assumed to be terminated.

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

Overall Conservatisms As described above, each of the changes made to the LOCA Analysis listed in Table 3-1 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, a holistic review of the model changes was performed to confirm no inadvertent nonconservatisms have been introduced. The results conclude that the model remains sufficiently conservative and consistent with the prescriptive requirements of RG 1.183.

RG 1.183 Appendix A Section 3.3 is prescriptive with respect to drywell spray removal and this license amendment request demonstrates compliance with these requirements, but per RG 1.183 Appendix A Section 4.5, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis. The methodology associated with modeling aerosol removal within the MSLs is unchanged from the analysis approved in Reference 6.2 except that the revised analysis replaces removal due to natural deposition with removal due to drywell spray. Therefore, it is important to determine the impact of crediting these removal methods concurrently.

NUREG/CR-0009 (Reference 6.15) is a compilation of experimental and theoretical information used by the NRC to develop the accident analysis spray removal methodology. This report primarily is based on the containment systems experimental data described in report BNWL-1457 (Reference 6.16). Per NUREG/CR-0009, aerosol removal by containment sprays Page 15

ATTACHMENT 1 Evaluation of Proposed Changes 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 MSIV 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 analysis are based on the conservative values in Section 6.5.2 of NUREG-0800. 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 6.17) 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 approved main steam line aerosol removal model does not include deposition by thermophoresis, diffusiophoresis, or flow irregularities.

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

Summary The revised LOCA dose consequence analysis performed in support of this license amendment request is based on the AST analysis previously approved by the NRC in Reference 6.2. The revised analysis 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). Since the MSIVs are postulated to leak at a total design leak rate of 250 scfh for Unit 1 and 350 scfh for Unit 2, Table 3-2 provides the bounding results (i.e.,

Unit 1 and 2 results are mixed together). EQ doses are not impacted by the changes associated with this LAR because the current EQ design basis does not include source term in Page 16

ATTACHMENT 1 Evaluation of Proposed Changes 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.

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

Activity Release Receptor Location Path Control Room EAB LPZ Containment Leakage 2.02E-01 2.88E-01 4.36E-01 ESF Leakage 8.95E-03 5.37E-03 9.90E-02 MSIV Leakage 2.92E+00 1.66E+01 2.94E+00 Reactor Building Shine 9.21E-02 0.00E+00 0.00E+00 External Cloud Shine 3.59E-01 0.00E+00 0.00E+00 CR Filter Shine negligible 0.00E+00 0.00E+00 Total 3.58E+00 1.69E+01 3.47E+00 CLB Doses 4.07E+00 8.85E+00 2.45E+00 Allowable TEDE Limit 5.00E+00 2.50E+01 2.50E+01 The control room dose decreased as compared to the CLB value primarily due to crediting drywell sprays to reduce isotopes escaping containment. The control room dose corresponds to a design leakage of 250 scfh from Unit 1. The EAB and LPZ doses correspond to a design leakage of 350 scfh from Unit 2. The control room doses from a release from Unit 2 are lower than the Unit 1 doses due to crediting the lower /Q value from a Unit 2 release from the MSIVs to the control room.

3.2 New TS 3.6.2.6 Residual Heat Removal (RHR) 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. The containment cooling mode of RHR is a safety function and consists of two cooling functions, containment spray which consists of drywell spray and suppression chamber spray along with suppression pool cooling. All containment cooling functions are manually initiated and all equipment and piping in the RHR system that feeds the containment spray nozzles are safety related. In addition, the containment spray system can be utilized for pressure suppression and heat removal. Because the drywell spray function now meets the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii),

the surveillance requirements are being moved from the technical requirements manual to the technical specifications.

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ATTACHMENT 1 Evaluation of Proposed Changes 3.3 Applicability of TSTF-551 Safety Evaluation Prior to this license amendment request, the QCNPS, Units 1 and 2 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 6.5), 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 conservatively bound the summer 1% and winter 99% values of 90F and -3F, respectively, for the Quad Cities area (i.e. Moline, Illinois) from Reference 6.13, Chapter 26 Tables 1A and 1B.

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 33 ft and 196 ft, which bound the height of the secondary containment. The assumed wind speed is conservative compared to the 5% wind speed of 20 mph for Moline, Illinois from Reference 6.13.

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 primary SGT System fan is loaded onto the diesel generator (DG) bus at 40 seconds after the LOCA occurs.

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ATTACHMENT 1 Evaluation of Proposed Changes

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 25 second time delay.

6. The SGT System flow rate is controlled to a maximum of 4000 cfm after the SGT System isolation valves open.

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 longer drawdown times than the summer cases (Cases 1 and 2) due to the higher outside air densities, which also tend to decrease the outside are pressure on the downwind side. The limiting drawdown time for the winter conditions with wind (Case 4) is 1,376 seconds. Therefore, the design basis secondary containment drawdown time is 1,376 seconds, or approximately 23 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.

QCNPS has confirmed 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 23 minute (1,376 seconds) 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 Quad Cities. 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 QCNPS, Units 1 and 2 and justify this amendment for the incorporation of the changes to the QCNPS, Units 1 and 2 TS.

The QCNPS Units 1 and 2 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). QCNPS, Units 1 and 2 were not licensed to the 10 CFR 50, Appendix A, GDC. The QCNPS, Units 1 and 2 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 Page 19

ATTACHMENT 1 Evaluation of Proposed Changes GDC. This difference does not alter the conclusion that the proposed change is applicable to QCNPS, Units 1 and 2.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The revision to the loss of coolant accident (LOCA) radiological dose analysis is performed using the Alternative Source Term 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 Population Zone (LPZ) are all below the regulatory dose limits.

Criterion 3 of 10 CFR 50.36(c)(2)(ii) states that Technical Specifications (TS) Limiting Conditions for Operation (LCO) must be established for structures, systems, or components (SSCs) that are part of the primary success path and which function or actuate to mitigate a design basis accident (DBA) or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into TS only those SSCs that are part of the primary success path of a safety sequence analysis.

Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria.

The revised LOCA dose consequence analysis credits the drywell sprays as part of the primary success path for accident or transient mitigation. Therefore, the addition of LCO 3.6.2.6, Residual Heat Removal (RHR) Drywell Spray, is part of the success path which functions to mitigate the consequences of a LOCA.

Per 10 CFR 50.36(c)(3), "Surveillance requirements 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 do not alter the design of secondary containment or its ability to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity. Therefore, the proposed revision to Surveillance Requirement (SR) 3.6.4.1.1 does not affect compliance with this regulation.

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.

Page 20

ATTACHMENT 1 Evaluation of Proposed Changes 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, "Residual Heat Removal (RHR) 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, EGC requests adoption of TSTF-551, "Revise Secondary Containment Surveillance Requirements," which is an approved change to the Standard Technical Specifications (STS),

into the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 Accidents at Nuclear Power Reactors" 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. As a result, no new failure modes of the MSIVs have been introduced.

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 Page 21

ATTACHMENT 1 Evaluation of Proposed Changes using the plant licensing basis for this accident. The resulting doses are slightly higher than the previously approved AST doses; with exception of the Control Room dose that is slightly lower.

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

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 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.

Page 22

ATTACHMENT 1 Evaluation of Proposed Changes The analysis has been performed using conservative methodologies. Safety margins and analytical conservatisms have been evaluated and have been 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 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 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 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)

Page 23

ATTACHMENT 1 Evaluation of Proposed Changes 6.2. Letter from Maitri Banerjee (U.S. NRC) to Christopher M. Crane, 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) 6.3. 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 (EPIDL 2017 LLA 0379)," dated January 9, 2018 (ADAMS Accession No. ML17353A949) 6.4. 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 (ADAMS Accession No. ML18024B022) 6.5. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792) 6.6. 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) 6.7. Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specifications 6.8. U.S. NRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC Generic Letter 99-02, dated June 3, 1999 and corresponding Errata dated August 23, 1999 6.9. NUREG-0800, Standard Review Plan Section 6.5.2, March 2007 6.10. Quad Cities Nuclear Power Station Technical Requirements Manual 3.6.a, Residual Heat Removal (RHR) Drywell Spray, Revision 0 6.11. J.E. Cline & Associates, Inc., "MSIV Leakage Iodine Transport Analysis," Letter Report dated March 26, 1991 (ADAMS Accession No. ML003683718) 6.12. Quad Cities Nuclear Power Station Updated Final Safety Analysis Report (UFSAR),

Revision 14 6.13. Fundamentals Handbook, ASHRAE, 1997 6.14. 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)

Page 24

ATTACHMENT 1 Evaluation of Proposed Changes 6.15. 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 6.16. BNWL-1457, Natural Transport Effects on Fission Product Behavior in the Containment Systems Experiment, R. K. Hilliard and L. F. Coleman, December 1970 6.17. WASH-1400 (NUREG-75/014), An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants Appendices VII, VIII, IX, and X, October 1975 (ADAMS Accession No. ML070600376) 6.18. QCTP 0130-01, Revision 26, Leak Rate Testing Program 6.19. 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 February 2019 from https://digital.library.unt.edu/ark:/67531/metadc101034/m2/1/high_res_d/metadc101034.

pdf Page 25

ATTACHMENT 2 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 MARK-UP OF QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS PAGES

PCIVs 62.4 scfh for Unit 1 3.6.1.3 and 78 scfh for Unit 2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is 34 scfh when tested at with the 25 psig, and the combined leakage rate Primary for all MSIV leakage paths is 86 scfh Containment when tested at 25 psig. Leakage Rate Testing Program 156 scfh for Unit 1 and 218 scfh for Unit 2 Quad Cities 1 and 2 3.6.1.3-8 Amendment No. 233/229

RHR Drywell Spray 3.6.2.6 3.6 CONTAINMENT SYSTEMS 3.6.2.6 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.6 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

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

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

C. Required Action and -------------NOTE------------

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.6.2.6-1 Amendment No.

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

SR 3.6.2.6.2 Verify each drywell spray nozzle is In accordance unobstructed. with the Surveillance Frequency Control Program SR 3.6.2.6.3 Verify RHR drywell spray subsystem In accordance locations susceptible to gas accumulation with the are sufficiently filled with water. Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.2.6-2 Amendment No.

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is In accordance 0.10 inch of vacuum water gauge. with the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify one secondary containment access In accordance door in each access opening is closed, with the except when the access opening is being Surveillance used for entry and exit. Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be In accordance maintained 0.25 inch of vacuum water with the gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem Surveillance at a flow rate 4000 cfm. Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment In accordance equipment hatches are closed and sealed. with the Surveillance Frequency Control Program

---

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

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

Quad Cities 1 and 2 3.6.4.1-2 Amendment No. 265/2609

ATTACHMENT 3 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 CLEAN QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS PAGES

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is 62.4 scfh for Unit 1 and with the 78 scfh for Unit 2 when tested at Primary 25 psig, and the combined leakage rate Containment for all MSIV leakage paths is 156 scfh Leakage Rate for Unit 1 and 218 scfh for Unit 2 when Testing Program tested at 25 psig.

Quad Cities 1 and 2 3.6.1.3-8 Amendment No. 233/229

RHR Drywell Spray 3.6.2.6 3.6 CONTAINMENT SYSTEMS 3.6.2.6 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.6 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

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

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

C. Required Action and -------------NOTE------------

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.6.2.6-1 Amendment No.

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

SR 3.6.2.6.2 Verify each drywell spray nozzle is In accordance unobstructed. with the Surveillance Frequency Control Program SR 3.6.2.6.3 Verify RHR drywell spray subsystem In accordance locations susceptible to gas accumulation with the are sufficiently filled with water. Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.2.6-2 Amendment No.

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

Verify secondary containment vacuum is 0.10 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify one secondary containment access In accordance door in each access opening is closed, with the except when the access opening is being Surveillance used for entry and exit. Frequency Control Program SR 3.6.4.1.3 Verify the secondary containment can be In accordance maintained 0.25 inch of vacuum water with the gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem Surveillance at a flow rate 4000 cfm. Frequency Control Program SR 3.6.4.1.4 Verify all secondary containment In accordance equipment hatches are closed and sealed. with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.6.4.1-2 Amendment No. 265/2609

ATTACHMENT 4 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 MARK-UP OF QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS BASES PAGES (For Information Only)

Primary Containment B 3.6.1.1 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 during the first 24 containment (La) is 3.0% by weight of the containment air hours following a per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a containment pressure of 43.9 psig.

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

LCO Primary containment OPERABILITY is maintained by limiting atmospheric 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA. MSIV leakage is not considered part of La Quad Cities 1 and 2 B 3.6.1.1-2 Revision 42

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 REQUIREMENTS (continued) This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) are OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break condition. This test is performed by blowing down the instrument line during an inservice leak or hydrostatic test and verifying a distinctive "click" when the poppet valve seats or a quick reduction in flow. This SR provides assurance that the instrumentation line EFCVs will perform as designed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.1.3.9 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 156 scfh for Unit 1 and each main steam isolation valve path is assumed to be the 218 scfh for Unit 2 minimum pathway leakage (smaller of either the inboard or outboard isolation valves individual leakage rates). The combined leakage rate limit through all MSIV leakage paths must be < 86 scfh when tested at > 25 psig for both as-left 250 scfh for Unit 1 and and as-found leakage rate tests. Additionally, the leakage 350 scfh for Unit 2 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 (continued)

Quad Cities 1 and 2 B 3.6.1.3-14 Revision 43 62.4 for Unit 1 and 78 for Unit 2

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.10 (continued) 43.9 of 60 scfh, when tested at 48 psig. This ensures that MSIV 100 scfh for Unit 1 leakage is properly accounted for in determining the overall and 125 scfh for impacts of primary containment leakage. The Frequency is Unit 2 required by the Primary Containment Leakage Rate Testing Program.

MSIV leakage is considered part of La.

REFERENCES 1. Technical Requirements Manual.

2 UFSAR, Section 15.6.5.

3. UFSAR, Section 15.6.4.

4. UFSAR, Chapter 15.

5. UFSAR, Section 5.2.2.2.3.

6. UFSAR, Section 6.2.4.1.

Quad Cities 1 and 2 B 3.6.1.3-15 Revision 43

RHR Drywell Spray B 3.6.2.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.6 Residual Heat Removal (RHR) Drywell Spray BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR drywell spray system provides overpressure protection to the primary containment by quenching steam released to the drywell during a loss of coolant accident (LOCA). Containment cooling is achieved by operating one RHR loop in the containment spray mode. This function is provided by two redundant drywell spray subsystems. During drywell spray operation, water pumped through the RHR heat exchangers would be diverted to spray headers in the drywell and above the suppression pool. The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES.

Each of the two RHR drywell spray subsystems contains two pumps, one heat exchanger, drywell spray valves, and a spray header in the drywell. Each RHR drywell spray subsystem is capable of recirculating water from the RHR suppression pool through a heat exchanger and the RHR drywell spray nozzles.

The spray then effects a temperature and pressure reduction through the combined effects of evaporative and convective cooling depending on the drywell atmosphere.

The drywell spray is also operated post-LOCA to wash, or scrub, inorganic iodines and particulates from the drywell atmosphere into the suppression pool.

The drywell spray can be used post-LOCA for both the scrubbing function as well as the temperature and pressure reduction effects. The radiological dose analysis does take credit of the RHR drywell spray system for scrubbing radionuclides for the drywell air space. Drywell spray is not required to maintain the drywell temperatures and pressures below the design limits.

The drywell spray mode of RHR is described in the FSAR, Reference 1.

(continued)

Quad Cities 1 and 2 B 3.6.2.6-1 Revision XX

RHR Drywell Spray B 3.6.2.6 BASES (continued)

APPLICABLE The RHR 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. The RHR drywell spray can also be used to reduce the temperature and pressure in the drywell.

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

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

LCO In the event of a LOCA, a minimum of one RHR drywell spray subsystem using one RHR pump is required to adequately scrub the inorganic iodines and particulates from the primary containment atmosphere. To ensure that these requirements are met, two RHR 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. An RHR drywell spray subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.

Management of gas voids is important to RHR drywell spray system OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause pressurization of, and the release of 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 RHR drywell spray subsystems OPERABLE is not required in MODE 4 or 5.

ACTIONS A.1 With one RHR drywell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE RHR drywell spray subsystem is adequate to perform the (continued)

Quad Cities 1 and 2 B 3.6.2.6-2 Revision XX

RHR Drywell Spray B 3.6.2.6 BASES ACTIONS A.1 (continued) primary containment fission product scrubbing and temperature and pressure reduction functions. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in the loss of the scrubbing and temperature and pressure reduction capabilities of the RHR drywell spray system. The 7 day Completion Time was chosen in light of the redundant RHR drywell spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.

B.1 With both RHR drywell spray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this condition, there is a substantial loss of the fission product scrubbing and temperature and pressure reduction functions of the RHR drywell spray system. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA.

C.1 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action C.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation (continued)

Quad Cities 1 and 2 B 3.6.2.6-3 Revision XX

RHR Drywell Spray B 3.6.2.6 BASES ACTIONS C.1 (continued) in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.

However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. 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 RHR 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 RHR 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.

SR 3.6.2.6.2 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.

(continued)

Quad Cities 1 and 2 B 3.6.2.6-4 Revision XX

RHR Drywell Spray B 3.6.2.6 BASES SURVEILLANCE SR 3.6.2.6.3 REQUIREMENTS (continued) RHR drywell spray system 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 RHR drywell spray subsystems and may also prevent water hammer and pump cavitation.

Selection of RHR drywell spray system 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 RHR drywell spray system 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 RHR drywell spray system 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.

RHR drywell spray system 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 (continued)

Quad Cities 1 and 2 B 3.6.2.6-5 Revision XX

RHR Drywell Spray B 3.6.2.6 BASES SURVEILLANCE SR 3.6.2.6.3 (continued)

REQUIREMENTS (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 Quad Cities 1 and 2 B 3.6.2.6-6 Revision XX

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1 and C.2 (continued)

Movement of recently irradiated fuel assemblies in the secondary containment and OPDRVs 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.

Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit is based on the expected short duration of the situations when the Note would be applied.

Quad Cities 1 and 2 B 3.6.4.1-4 Revision 43

Secondary Containment B 3.6.4.1 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at a flow rate 4000 cfm.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 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. UFSAR, Section 9.1.4.3.2.

3. NRC Safety Evaluation Report for the Holtec International HI-STORM 100 Storage System (Docket Number 72-1014, Certificate Number 1014, Amendment 2).

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

5. QDC-7500-M-2341, Revision 0, Quad Cities Units 1

& 2 Secondary Containment Drawdown Analysis, February 2019.

Quad Cities 1 and 2 B 3.6.4.1-6 Revision 54