RS-23-007, Application to Adopt TSTF-564, Safety Limit MCPR

Application to Adopt TSTF-564, Safety Limit MCPR

ML23062A450
Person / Time
Site:	Quad Cities
Issue date:	03/03/2023
From:	Simpson P Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-23-007
Download: ML23062A450 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.90 RS-23-007 March 3, 2023 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 Adopt TSTF-564, "Safety Limit MCPR"

References:

1. TSTF-564, Revision 2, "Safety Limit MCPR," dated October 24, 2018 (ML18297A361)

2. Final Safety Evaluations of Technical Specifications Task Force Traveler TSTF-564, Revision 2, "Safety Limit MCPR," dated November 16, 2018 (ML18299A048)

3. TSTF-22-09, "Notification of an Error in Approved Traveler TSTF-564, Safety Limit MCPR," dated October 14, 2022 (ML22290A167)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 respectively.

CEG requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2 (References 1 and 2),

which is an approved change to the Improved Standard Technical Specifications (ISTS), into the station's TS. The proposed change revises the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked-up to show the proposed TS changes. Attachment 3 provides TS Bases pages marked up to show the associated TS Bases changes and is provided for information only.

March 3, 2023 U.S. Nuclear Regulatory Commission Page 2 The QCNPS Plant Operations Review Committee has reviewed the proposed change in accordance with the requirements of the CEG Quality Assurance Program.

CEG requests approval of the proposed license amendment by March 1, 2024, to support QCNPS implementation prior to startup from the spring 2024 Unit 2 refueling outage.

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 3rd day of March 2023.

Respectfully, Patrick R. Simpson Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Changes

2. Proposed Mark-up of Quad Cities, Units 1 and 2 Technical Specifications Pages

3. Mark-up of Quad Cities, Units 1 and 2 Technical Specifications Bases Pages - For Information Only cc: NRC Regional Administrator, Region III 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 Adopt TSTF-564, "Safety Limit MCPR" 1.0

SUMMARY

DESCRIPTION

2.0 TECHNICAL EVALUATION

2.1 Applicability of Safety Evaluation 2.2 Variations 2.3 Detailed Description of the Change

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements/Criteria 3.2 Precedents 3.3 No Significant Hazards Consideration 3.4 Conclusion

5.0 REFERENCES

Page 1 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION Constellation Energy Generation, LLC (CEG) requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2 (Reference 5.1), which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 Technical Specifications (TS). The proposed amendment revises the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.

2.0 TECHNICAL EVALUATION

2.1 Applicability of the Safety Evaluation CEG has reviewed the safety evaluation for TSTF-564 (Reference 5.1) provided to the Technical Specifications Task Force (TSTF) in a letter dated November 16, 2018. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. CEG has concluded that the changes are applicable to QCNPS and incorporation of these changes into the QCNPS Technical Specifications (TS) is justified.

The QCNPS Units 1 and 2 reactors are currently fueled with Framatome ATRIUM-10XM fuel bundles, but CEG plans to begin loading Global Nuclear Fuels Americas, LLC (GNF) GNF3 fuel into the QCNPS Unit 1 core beginning with spring 2023 refueling outage, with the Unit 2 transition beginning with the spring 2024 refueling outage.

CEG has reviewed TSTF-22-09 (Reference 5.2) and confirmed that the GNF equation error identified in that document has no impact on the adoption of the Traveler by QCNPS.

As described in TSTF-564, for transition cores loaded with a mix of fuel types, the SLMCPR95/95 is based on the largest MCPR95/95 value for the fuel types used. At the time of initial implementation at each unit, the core will contain both once-burned ATRIUM 10XM and fresh GNF3 fuel assemblies. Framatome has calculated a Safety Limit value of 1.05 for the ATRIUM-10XM fuel type using Equation 1 in Section 3.1 of TSTF-564 (References 5.3, 5.4, and 5.6). The corresponding value for GNF3 from TSTF-564 Table 1 is 1.07. The proposed Safety Limit in TS 2.1.1.2 is 1.07, consistent with the TSTF-564 guidance to use the largest MCPR95/95 value for the transition cores containing both ATRIUM-10XM and GNF3. TS Bases mark-up provided in Attachment 3 identified GNF3 as the fuel type the SL is based upon.

The MCPR value calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR99.9%. Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," subsection a.2 is revised to require the MCPR (existing requirement) and MCPR99.9% (new requirement) values to be included in the cycle-specific COLR.

2.2 Variations CEG is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation dated November 16, 2018.

Page 2 of 7

ATTACHMENT 1 Evaluation of Proposed Changes The QCNPS TS utilize different numbering and titles than the Standard Technical Specifications on which TSTF-564 was based. Specifically, Section 5.6.5, "Core Operating Limits Report (COLR)" is numbered 5.6.3 in TSTF-564. This difference is administrative and does not affect the applicability of TSTF-564 to the QCNPS TS.

The QCNPS TS contain requirements that differ from the Standard TS on which TSTF-564 was based, such as reactor steam dome pressure in TS 2.1.1.2 and Applicability in TS 3.2.2, but these differences do not affect the applicability of the TSTF-564 justification.

Additionally, QCNPS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). QCNPS UFSAR Section 3.1 "Conformance with NRC Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concludes that the plant specific requirements are sufficiently similar to the Appendix A GDC. Therefore, this difference does not alter the conclusion that the proposed change is applicable to QCNPS.

2.3 Detailed Description of Change A single SLMCPR value will be used in TS 2.1.1.2 instead of two values applicable when one or two recirculation loops are in operation. Additionally, TS 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the MCPR (existing requirement) and MCPR99.9% (new requirement) values to be included in the cycle-specific COLR. contains a marked-up version of the TS showing the proposed changes. provides the marked-up TS Bases pages. The TS Bases mark-up pages are being submitted for information only.

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36. The applicant must also include in the application a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls. . . ." However, per 10 CFR 50.36(a)(1), these TS bases "shall not become part of the technical specifications."

As required by 10 CFR 50.36I(1), the QCNPS TS include safety limits, limiting safety system settings, and limiting control settings. The regulation, 10 CFR 50.36(c)(1)(i)(A), states, in part:

Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.

Page 3 of 7

ATTACHMENT 1 Evaluation of Proposed Changes The limits placed on the MCPR act as an acceptable fuel design limit to prevent boiling transition, which has the potential to result in fuel rod cladding failure, which meets General Design Criteria (GDC) 10, "Reactor Design" in 10 CFR 50, Appendix A, which states in part:

The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The original design of QCNPS, Units 1 and 2 was reviewed and approved against the draft GDC issued in July 1967. Updated Final Safety Analysis Report (UFSAR), Section 3.1, "Conformance with NRC General Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concluded that the plant specific requirements are sufficiently similar to the Appendix A GDC.

3.2 Precedents Per Reference 5.2, all U.S. BWR plants except Dresden Units 2 and 3, Nine Mile Point Unit 1, and Quad Cities Units 1 and 2 have adopted or submitted an amendment to adopt TSTF-564.

The other BWRs in the CEG fleet (Clinton, FitzPatrick, LaSalle, Limerick, Nine Mile Point 2, and Peach Bottom) previously adopted TSTF-564 per Reference 5.5 using a GNF fuel product listed in Table 1 of TSTF-564. That request did not include Quad Cities because both units utilized Framatome ATRIUM 10XM fuel, which is not included in Table 1 of TSTF-564. Since the approval of the CEG fleet submittal, Framatome has calculated a MCPR95/95 value for ATRIUM 10XM using a method consistent with TSTF-564. Brunswick (Reference 5.3),

Monticello (Reference 5.4), and Susquehanna (Reference 5.6) have received approval for adopting TSTF-564 using the Framatome developed MCPR95/95 for ATRIUM 10XM.

3.3 No Significant Hazards Consideration Overview CEG requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the QCNPS, Units 1 and 2 Technical Specifications (TS). The proposed change revises the TS safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the MCPR (existing requirement) and MCPR99.9% (new requirement) values to be included in the COLR.

CEG 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:

Page 4 of 7

ATTACHMENT 1 Evaluation of Proposed Changes

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

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling.

The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated.

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 proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems, or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

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 The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95% probability with 95% confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9%

of the fuel rods are not susceptible to boiling transition does not have a significant effect on the plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.

Page 5 of 7

ATTACHMENT 1 Evaluation of Proposed Changes Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.4 Conclusion In 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.

4.0 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 change.

5.0 REFERENCES

5.1 TSTF-564, Revision 2, "Safety Limit MCPR," (ML18297A361) dated October 24, 2018 and associated NRC Safety Evaluation dated November 16, 2018 (ML18299A048) 5.2 TSTF-22-09, "Notification of an Error in Approved Traveler TSTF-564, Safety Limit MCPR," dated October 14, 2022 (ML22290A167) 5.3 Letter from A. Hon (U.S. NRC) to W.R. Gideon (Brunswick Steam Electric Plant),

"Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendments to Revise Technical Specification Section 2.1.1.2, Safety Limit Minimum Critical Power Ratio (CAC Nos. MF8470 and MF8741)," dated March 10, 2017 (ML17059D146) 5.4 Letter from R.F. Kuntz (U.S. NRC) to T.A. Conboy (Northern States Power Company -

Minnesota), "Monticello Nuclear Generating Plant - Issuance of Amendment No. 207 RE: Adoption of TSTF-564, Safety Limit MCPR (EPID L-2020-LLA-0243)," dated October 15, 2021 (ML21223A280)

Page 6 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 5.5 Letter from B.A. Purnell (U.S. NRC) to B.C. Hanson (Exelon Generation Company, LLC),

"Clinton Power Station, Unit No. 1; James A. Fitzpatrick Nuclear Power Plant; Lasalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Nine Mile Point Nuclear Station, Unit 2; and Peach Bottom Atomic Power Station, Units 2 and 3 -

Issuance of Amendments to Adopt TSTF-564, 'Safety Limit MCPR' (EPID L-2019-LLA-0021)," dated August 28, 2019 (ML19176A033) 5.6 Letter from A.L. Klett (U.S. NRC) to K. Cimorelli (Talen Energy), "Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Nos. 281 and 264 Re: Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR (EPID L-2022-LLA-0005)," dated July 15, 2022 (ML22146A207)

Page 7 of 7

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

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 685 psig or core flow < 10% rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 685 psig and core flow 10% rated core flow:

For two recirculation loop operation, MCPR shall be 1.08, or for single recirculation loop operation, MCPR shall be 1.10.

1.07 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1345 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

Quad Cities 1 and 2 2.0-1 Amendment No. 276/271

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

(ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Report

---

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

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 (Deleted) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The APLHGR for Specification 3.2.1.

2. The MCPR for Specification 3.2.2.

and MCPR99.9% (continued)

Quad Cities 1 and 2 5.6-2 Amendment No. 225/220

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 MARK-UP OF QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS BASES PAGES (For Information Only)

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND UFSAR Section 3.1.2.1 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.

(continued)

This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.

Quad Cities 1 and 2 B 2.1.1-1 Revision 0

Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation. Establishment of Emergency Core Cooling System initiation setpoints higher than this SL provides margin such that the SL will not be reached or exceeded.

APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

The Technical The Reactor Protection System setpoints (LCO 3.3.1.1, Specifications SL is set "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any generically on a fuel anticipated combination of transient conditions for Reactor product MCPR Coolant System water level, pressure, and THERMAL POWER correlation basis as the level that would result in reaching the Safety Limit MCPR.

MCPR which corresponds to a 95% Cores with fuel that is all from one vendor utilize that probability at a 95% vendor's critical power correlation for determination of MCPR. For cores with fuel from more than one vendor, the confidence level that MCPR is calculated for all fuel in the core using the transition boiling will not licensed critical power correlations. This may be occur, referred to as accomplished by using each vendor's correlation for the SLMCPR95/95. vendor's respective fuel. Alternatively, a single correlation can be used for all fuel in the core. For fuel that has not been manufactured by the vendor supplying the critical power correlation, the input parameters to the reload vendor's correlation are adjusted using benchmarking data to yield conservative results compared with the critical power results from the co-resident fuel.

(continued)

Quad Cities 1 and 2 B 2.1.1-2 Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity SAFETY ANALYSES (continued) The AREVA ACE/ATRIUM 10XM correlation is valid for critical power calculations at pressures > 300 psia and at bundle mass fluxes > 0.1 x 106 lb/hr-ft2 (Ref. 2). The GNF GEXL98 correlation is applicable to ATRIUM 10XM fuel monitored with the GNF core monitoring system. The GEXL98 correlation is valid for critical power calculations at pressures and bundle mass fluxes consistent with the range of 6 applicability for the ACE/ATRIUM 10XM correlation (Ref.8).

The GNF GEXL21 correlation is applicable to GNF3 fuel and is valid for critical power calculations at pressures 600 psia and core flows 10% of rated flow (Ref. 9).

7 For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi.

Analyses show that with a bundle flow of 28 x 103 lb/hr (approximately a mass velocity of 0.25 X 106 lb/hr-ft2), bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale critical power test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 685 psig is conservative. Although the ACE/ATRIUM 10XM, GEXL21, and GEXL98 correlations are valid at reactor steam dome pressures > 600 psia, application of the fuel cladding integrity SL at reactor steam dome pressure

< 685 psig is conservative. Additional information on low flow conditions is available in Reference 4.

(continued)

Quad Cities 1 and 2 B 2.1.1-3 Revision 79

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The SLMCPR ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the SLMCPR is based on a detailed statistical procedure that considers the uncertainties in For cores with a single fuel monitoring the core operating state. One specific product line, the SLMCPR95/95 uncertainty included in the SL is the uncertainty inherent is the MCPR95/95 for the fuel in the fuel vendor's critical power correlation.

References 2, 3, 5, 7, 8 and 9 describe the methodologies type. For cores loaded with a used in determining the SLMCPR.

mix of applicable fuel types, the SLMCPR95/95 is based on The fuel vendors critical power correlation is based on a the largest (i.e., most limiting) significant body of practical test data, providing a high of the MCPR values for the degree of assurance that the critical power, as evaluated by fuel product lines that are fresh the correlation, is within a small percentage of the actual or once-burnt at the start of the critical power being estimated. As long as the core cycle. In cores containing both pressure and flow are within the range of validity of the ATRIUM 10XM and GNF3 fuel, correlation, the assumed reactor conditions used in defining the SLMCPR is based on the the SL introduce conservatism into the limit because more limiting GNF3 fuel. bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the inherent accuracy of the fuel vendor's correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the SLMCPR. If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

(continued)

The SLMCPR is set such that no significant fuel damage is calculated to occur if the limit is not violated.

Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent.

The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95%

confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.

Quad Cities 1 and 2 B 2.1.1-4 Revision 79

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level SAFETY ANALYSES (continued) During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 2.2 5 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 6). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

(continued)

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Reactor Core SLs B 2.1.1 BASES (continued)

REFERENCES 1. UFSAR, Section 3.1.2.1.

2. ANP-10298P-A Revision 1, "ACE/ATRIUM 1OXM Critical Power Correlation," March 2014 (As specified in Technical Specification 5.6.5).

3. NEDE-24011-P-A, "General Standard Application for Reactor Fuel" (As specified in Technical Specification 5.6.5).

4. General Electric Services Information Letter (SIL)

No. 516, Supplement 2, January 19, 1996.

5. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011 (As specified in Technical Specification 5.6.5).

6.

5. 10 CFR 50.67.

7. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000 (As specified in Technical Specification 5.6.5).

8.

6. NEDC-33930P Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021 (As specified in Technical Specifications 5.6.5.).

9.

7. NEDC-33880P, Revision 1, "GEXL21 Correlation for GNF3 Fuel," November 2017.

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MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The Safety Limit MCPR (SLMCPR) is set such that 99.9% of the fuel rods are expected to avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which

, and that 99.9% of boiling transition is calculated to occur has been adopted the fuel rods are as a fuel design criterion.

not susceptible to boiling transition if The onset of transition boiling is a phenomenon that is the limit is not readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations violated. have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the AOOs to establish the operating limit MCPR are presented in References 2, 3, 4, 5, and 6. To ensure that the SLMCPR is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting combined with the MCPR99.9% transient yields the largest change in CPR (CPR). When the largest CPR is added to the SLMCPR, the required operating limit MCPR is obtained. MCPR Insert 1 are 99.9%

The MCPR operating limits derived from the transient value and the analysis are dependent on the operating core flow state

, and (continued)

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INSERT 1 MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. MCPR99.9% is expected to always be greater than MCPR95/95 because MCPR99.9% includes uncertainties not factored into MCPR95/95 and because the 99.9 probability basis for determining MCPR99.9% is more conservative than the 95 percent probability at a 95 percent confidence level used in determining the MCPR95/95. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPR99.9% calculation are given in Reference 6. Reference 6 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9% statistical analysis.

MCPR B 3.2.2 BASES and MCPRp, respectively APPLICABLE (MCPRf) to ensure adherence to fuel design limits during the SAFETY ANALYSES worst transient that occurs with moderate frequency as (continued) identified in UFSAR, Chapter 15 (Ref. 4).

Flow-dependent MPCR limits, MCPRf, ensure that the SLMCPR is not violated during recirculation flow events. The design basis flow increase event is a slow-flow power increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow. Flow runout events are simulated along a constant xenon flow control line assuming a quasi steady-state plant heat balance. The MCPRf limit is specified as an absolute value and protects the SLMCPR. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

For GNF reloads, above the power at which the scram is bypassed (PBypass), bounding power-dependent trend functions have been developed. This trend function, Kp, is used as a multiplier to the rated MCPR operating limits to obtain the power-dependent MCPR limits, MCPRp. Below the power at which the scram is automatically bypassed (Below PBypass),

the MCPRP limits are actual absolute operating limit MCPR values, rather than multipliers on the rated power operating limit MCPR.

For Framatome reloads, the MCPRp limits are actual absolute operating limit MCPR values.

The power dependent limits are established to protect the SLMCPR.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the appropriate MCPRf or the rated condition MCPR limit.

(continued)

MCPRf and MCPRp limits, which are based on (MCPR99.9% value, the MCPR99.9% limit specified in the COLR. MCPRf values, and MCPRp values)

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