RS-23-071, Application to Adopt TSTF-564, Safety Limit MCPR

Application to Adopt TSTF-564, Safety Limit MCPR

ML23242A044
Person / Time
Site:	Dresden
Issue date:	08/30/2023
From:	Lueshen K Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-23-071
Download: ML23242A044 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-23-071 10 CFR 50.90 August 30, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Application to 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-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

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 DNPS, Units 2 and 3 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.

Attachment 1 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.

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

August 30, 2023 U.S. Nuclear Regulatory Commission Page 2 CEG requests approval of the proposed license amendment by August 30, 2024. The DNPS implementation will occur with the fall 2024 Unit 3 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 Mr. Mitchel Mathews at (779) 231-5765.

I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 30th day of August 2023.

Respectfully, Digitally signed by Lueshen, Lueshen, Kevin Kevin Date: 2023.08.30 07:29:04 -05'00' Kevin Lueshen Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Changes

2. Proposed Markup of Technical Specifications Pages

3. Proposed Markup of Technical Specifications Bases Pages (For Information Only) cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Dresden 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 Dresden Nuclear Power Station (DNPS), Units 2 and 3 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 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 DNPS and incorporation of these changes into the DNPS TS is justified.

The DNPS, Units 2 and 3, 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 DNPS, Unit 2 core beginning with the fall 2023 refueling outage, and DNPS, Unit 3 core beginning with the fall 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 DNPS.

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 implementation of this amendment, the DNPS, Units 2 and 3 cores 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 Technical Specification 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. GNF3 is identified in the TS Bases mark-up provided in 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.

Page 2 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 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.

The DNPS TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-564 was based. Specifically, COLR is Technical Specification 5.6.5 in the DNPS, Units 2 and 3 TS, and 5.6.3 in the applicable portion of TSTF-564. This difference is administrative and does not affect the applicability of TSTF-564 or the NRC staffs safety evaluation dated November 16, 2018, to the DNPS, Units 2 and 3 TS.

The DNPS TS contain requirements that differ from the STS on which TSTF-564 was based, such as reactor steam dome pressure in Technical Specification 2.1.1.2, but these differences do not affect the applicability of the TSTF-564 justification.

Additionally, DNPS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). DNPS Updated Final Safety Analysis Report (UFSAR), Section 3.1.1, "Compliance with Draft Design Criteria," provides an assessment against the 70 draft GDC published in 1967. The design basis of Unit 2 was later evaluated against the final "General Design Criteria for Nuclear Power Plants," published as 10 CFR 50, Appendix A in July 1971.

This evaluation is presented in UFSAR Section 3.1.2, "Compliance with Final Design Criteria,"

and concluded that DNPS, Unit 2 satisfies and is in compliance with the intent of the General Design Criteria. This evaluation was performed specifically for Unit 2 and may not fully apply to Unit 3; however, the high degree of similarity between the design of Units 2 and 3 indicates that Unit 3 also conforms to the intent of the General Design Criteria. Therefore, this difference does not alter the conclusion that the proposed change is applicable to DNPS.

2.3 Detailed Description of the Change A single SLMCPR value will be used in Technical Specification 2.1.1.2 instead of two values applicable when one or two recirculation loops are in operation. Additionally, Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the 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.

Page 3 of 7

ATTACHMENT 1 Evaluation of Proposed Changes

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 50.36(c)(1), the DNPS 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.

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.

A discussion of the applicability of the 10 CFR 50, Appendix A, GDC for DNPS, Units 2 and 3 is provided in Section 2.2 above.

3.2 Precedents Per Reference 5.2, all U.S. BWR plants except Dresden, Units 2 and 3, and Nine Mile Point, Unit 1, have adopted or submitted requests to adopt TSTF-564. CEG submitted an amendment request to adopt TSTF-564 for Quad Cities Nuclear Power Station, Units 1 and 2 on March 3, 2023 (Reference 5.7). The other BWRs in the CEG fleet (i.e., 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. Dresden and Quad Cities were not included in that request because they 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 Page 4 of 7

ATTACHMENT 1 Evaluation of Proposed Changes TSTF-564. Brunswick (Reference 5.3), Monticello (Reference 5.4), and Susquehanna (Reference 5.6) have received approval to adopt TSTF-564 using the Framatome-developed MCPR95/95 for ATRIUM 10XM.

3.3 No Significant Hazards Consideration Overview Constellation Energy Generation, LLC (CEG) requests to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Dresden Nuclear Power Station (DNPS), Units 2 and 3, 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:

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 change does 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 previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits Page 5 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 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 change does 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.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed amendment 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 Page 6 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 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) 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) 5.7 Letter form P.R. Simpson (Constellation Energy Generation, LLC), to U.S. NRC, "Application to Adopt TSTF-564, 'Safety Limit MCPR,'" dated March 3, 2023 (ML23062A450)

Page 7 of 7

ATTACHMENT 2 Proposed Markup of Technical Specifications Pages Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 TECHNICAL SPECIFICATIONS PAGES 2.0-1 5.6-2

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:

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

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.

Dresden 2 and 3 2.0-1 Amendment No. 263/256

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)

Dresden 2 and 3 5.6-2 Amendment No. 214/206

ATTACHMENT 3 Proposed Markup of Technical Specifications Bases Pages (For Information Only)

Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 TECHNICAL SPECIFICATIONS BASES PAGES B 2.1.1-1 B 2.1.1-2 B 2.1.1-3 B 2.1.1-4 B 2.1.1-6 B 2.1.1-7 B 3.2.2-1 B 3.2.2-2 INSERT

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

B 2.1.1 Reactor Core SLs BASES BACKGROUND UFSAR Section 3.1.2.2.1 (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 This is accomplished by source is incrementally cumulative and continuously having a Safety Limit measurable. Fuel cladding perforations, however, can Minimum Critical Power result from thermal stresses, which occur from reactor Ratio (SLMCPR) design operation significantly above design conditions.

basis, referred to as SLMCPR95/95, which While fission product migration from cladding perforation corresponds to a 95% is just as measurable as that from use related cracking, probability at a 95% the thermally caused cladding perforations signal a confidence level (the threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding 95/95 MCPR criterion) deterioration. Therefore, the fuel cladding SL is defined that transition boiling will with a margin to the conditions that would produce onset of not occur. 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)

Dresden 2 and 3 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 generically on a fuel any anticipated combination of transient conditions for product MCPR Reactor Coolant System water level, pressure, and THERMAL correlation basis as the POWER level that would result in reaching the Safety Limit MCPR which MCPR.

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 confidence level that MCPR. For cores with fuel from more than one vendor, the transition boiling will not MCPR is calculated for all fuel in the core using the occur, referred to as licensed critical power correlations. This may be SLMCPR95/95. accomplished by using each vendor's correlation for the 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)

Dresden 2 and 3 B 2.1.1-2 Revision 91

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 applicability for the ACE/ATRIUM 10XM correlation (Ref. 9). The GNF GEXL21 correlation is applicable to GNF3 7

fuel and is valid for critical power calculations at pressures 600 psia and core flows 10% of rated flow (Ref. 10).

8 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)

Dresden 2 and 3 B 2.1.1-3 Revision 91

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 SLMCPR is set such The margin between calculated boiling transition that no significant fuel (i.e., MCPR = 1.00) and the SLMCPR is based on a detailed damage is calculated to statistical procedure that considers the uncertainties in occur if the limit is not monitoring the core operating state. One specific violated. Since the uncertainty included in the SL is the uncertainty inherent parameters that result in in the fuel vendors critical power correlation.

fuel damage are not directly References 2, 3, 5, 7, 9, and 10 describe the methodologies observable during reactor used in determining the SLMCPR.

operation, the thermal and The fuel vendors critical power correlation is based on a hydraulic conditions that significant body of practical test data, providing a high result in the onset of degree of assurance that the critical power, as evaluated transition boiling have been by the correlation, is within a small percentage of the used to mark the beginning actual critical power being estimated. As long as the core of the region in which fuel pressure and flow are within the range of validity of the damage could occur. correlation, the assumed reactor conditions used in Although it is recognized defining the SL introduce conservatism into the limit that the onset of transition because bounding high radial power factors and bounding boiling would not result in flat local peaking distributions are used to estimate the damage to BWR fuel rods, number of rods in boiling transition. These conservatisms the critical power at which and the inherent accuracy of the fuel vendors correlation boiling transition is provide a reasonable degree of assurance that there would calculated to occur has be no transition boiling in the core during sustained been adopted as a operation at the SLMCPR. If boiling transition were to occur, there is reason to believe that the integrity of the convenient limit. The fuel would not be compromised. Significant test data Technical Specification SL accumulated by the NRC and private organizations indicate value is dependent on the that the use of a boiling transition limitation to protect fuel product line and the against cladding failure is a very conservative approach.

corresponding MCPR Much of the data indicate that BWR fuel can survive for an correlation, which is cycle extended period of time in an environment of boiling independent. The value is transition.

based on the Critical Power Ratio (CPR) data statistics (continued) and a 95% probability with For cores with a single fuel product line, the SLMCPR95/95 is the 95% confidence that rods MCPR95/95 for the fuel type. For cores loaded with a mix of applicable are not susceptible to fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of boiling transition, referred to the MCPR values for the fuel product lines that are fresh or once-burnt as MCPR95/95. at the start of the cycle. In cores containing both ATRIUM 10XM and GNF3 fuel, the SLMCPR is based on the more limiting GNF3 fuel.

Dresden 2 and 3 B 2.1.1-4 Revision 91

Reactor Core SLs B 2.1.1 BASES (continued)

SAFETY LIMIT 2.2 VIOLATIONS 5

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)

Dresden 2 and 3 B 2.1.1-6 Revision 91

Reactor Core SLs B 2.1.1 BASES (continued)

REFERENCES 1. UFSAR, Section 3.1.2.2.1.

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

3. NEDE-24011-P-A, "General Electric 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) 5.

6. 10 CFR 50.67.

7. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000. (As Specified in Technical Specification 5.6.5) 6.

8. UFSAR, Section 7.1.

7.

9. NEDC-33930P Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021. (As Specified in Technical Specification 5.6.5) 8.

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

Dresden 2 and 3 B 2.1.1-7 Revision 91

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

, and that 99.9% of experienced boiling transition (Ref. 1), the critical power the fuel rods are at which boiling transition is calculated to occur has been adopted 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 violated. designs. Based on these experimental data, correlations 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 combined with decrease. The limiting transient yields the largest change the MCPR99.9% in CPR (CPR). When the largest CPR is added to the SLMCPR, the required operating limit MCPR is obtained.

(continued)

INSERT Dresden 2 and 3 B 3.2.2-1 Revision 91

MCPR B 3.2.2 MCPR99.9%

BASES , and are value and the and power APPLICABLE The MCPR operating limits derived from the transient SAFETY ANALYSES analysis are dependent on the operating core flow state (continued) (MCPRf) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency as identified in UFSAR, Chapter 15 (Ref. 4).

and MCPRp, respectively 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 stead-state plant heat balance. The MCPRf limit is specified as an absolute value and is based on the MCPR safety limit.

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 safety limit.

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 the MCPR99.9% limit specified in the COLR. (MCPR99.9% value, MCPRf values, and MCPRp values)

Dresden 2 and 3 B 3.2.2-2 Revision 91

INSERT 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.