Text

Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)
	ML20063L295
Person / Time
Site:	Byron, Braidwood
Issue date:	02/28/2020
From:	Demetrius Murray; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-20-010
	Download: ML20063L295 (82)
	v • d • e

PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-20-010 10 CFR 50.90 February 28, 2020 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRG Docket Nos. 50-456 and 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRG Docket Nos. 50-454 and 50-455

Subject:

Application to Revise Technical Specifications 5.6.5, "Core Operating Limits Report (COLR)"

References:

1. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. NRG, "Response to Request for Information Regarding the Thermal Conductivity Degradation and 10 CFR 50.46 Report," dated March 19, 2012

2. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. NRG, "EGGS Evaluation Model Error- 10 CFR 50.46 Report," dated May 21, 2012

3. Letter from David M. Gullett (Exelon Generation Company, LLC) to U.S. NRG, "Supplement to RS-12-037 and RS-12-087: Revision to Exelon Generation Company, LLC Commitment Relating to Large Break Loss of Coolant Accident (LBLOCA) Analysis with an NRG Approved Emergency Core Cooling System (EGGS) Evaluation Model that Explicitly Accounts for Thermal Conductivity Degradation (TCD)," dated December 14, 2016

4. Letter from Dwi Murray (Exelon Generation Company, LLC) to U.S. NRG, "Supplement to RS-16-239: Revision to Exelon Generation Company, LLC Commitment Relating to Large Break Loss of Coolant Accident (LBLOCA)

Analysis with an NRG Approved Emergency Core Cooling System (EGGS)

Evaluation Model that Explicitly Accounts for Thermal Conducitivity Degradation (TCD)," dated September 5, 2019 Attachment 3 contains Proprietary Information.

When separated from Attachment 3, this document is decontrolled.

February 28, 2020 U.S. Nuclear Regulatory Commission Page 2 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is submitting a request for an amendment to the Technical Specifications (TS) for Renewed Facility Operating License Nos.

NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood), and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron).

This amendment request proposes to revise TS 5.6.5, "Core Operating Limits Report (COLR)"

The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL SPECTRUM' 1 LOCA Evaluation Model (FSLOCATM EM), that is contained in WCAP-16996-P-A, Rev. 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," and was used for LOCA reanalysis for Braidwood and Byron.

This License Amendment Request completes EGC's action to perform an ECCS LOCA reanalysis in a.ccordance with 10 CFR 50.46(a)(3)(ii) as described in Reference 1 and as

. amended in References 3 and 4.

Attachment .3 is a proprietary version of the Application of Westinghouse FULL SPECTRUM TM LOCA Evaluation Model to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2.

Attachment 3 contains information proprietary to Westinghouse Electric Company LLC.

Westinghouse Electric Company LLC and EGC request that the contents of Attachment 3 be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4). Attachment 4 provides the non-proprietary version of the Application of Westinghouse FULL SPECTRUM TM LOCA Evaluation Model to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2.

The attached request is subdivided as follows:

Attachment 1 provides a description and evaluation of the proposed changes.

Attachment 2 provides the Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCA TM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2."

Attachment 3 provides the Proprietary Class 2 Version of FSLOCATM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2."

Attachment 4 provides the Non-Proprietary Class 3 Version of the FSLOCATM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2."

Attachment Sa provides the markup of the affected TS pages for Braidwood.

1 FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, it subsidiaries and/or affiliates in the United States of America and may be registered in other countries through the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

February 28, 2020 U.S. Nuclear Regulatory Commission Page3

Attachment Sb provides the markup of the affected TS pages for Byron.

Attachment 6a provides the TS Bases pages marked to show the proposed changes for information only for Braidwood.

Attachment 6b provides the TS Bases pages marked to show the. proposed changes for information only for Byron.

The proposed change has been reviewed by Braidwood and Byron Plant Operations Review Committees in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed license amendment request within one year of this submittal date; i.e., by February 28, 2021. Once approved, the amendment shall be implemented according to the following table. It will be implemented during the defueled window in the specified Refueling Outage (Calendar Year (CY) provided for reference):

Unit Date Braidwood Unit 1 Refuelin B ran Unit 1 Refuelin Braidwood Unit 2 Refuelin B ran Unit 2 Refuelin In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

This letter contains no regulatory commitments. Should.you have any questions concerning this letter, please contact Ms. Lisa Zurawski at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of February 2020.

Dwi Murray Senior Manager - Licensing Exelon Generation Company, LLC

February 28, 2020 U.S. Nuclear Regulatory Commission Page4 Attachments:

1) Evaluation of Proposed Changes

2) Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCA' LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2"

3) Proprietary Class 2 Version of FSLOCA TM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2"

4) Non-Proprietary Class 3 Version of the FSLOCA TM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2" Sa) Proposed Technical Specification Changes (Mark-Up) for Braidwood Sb) Proposed Technical Specification Changes (Mark-Up) for Byron 6a) Proposed Technical Specifications Bases Changes (Mark-Up) for Information Only for Braidwood 6b) Proposed Technical Specifications Bases Changes (Mark-Up) for Information Only for Byron cc: NRG Regional Administrator, Region Ill NRG Senior Resident Inspector, Braidwood Station NRG Senior Resident Inspector, Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety

Attachment 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to revise Technical Specifications 5.6.5, "Core Operating Limits Report (COLR)"

1.0 DESCRIPTION

2.0 DETAILED DESCRIPTION 2.1 Proposed Changes

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Attachment 1 Evaluation of Proposed Changes

1.0 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 amendments to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, (Braidwood) and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron). This amendment request proposes to revise Technical Specifications (TS) 5.6.5, "Core Operating Limits Report (COLR)." The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL SPECTRUM' 2 LOCA Evaluation Model (FSLOCA' EM), that is contained in WCAP-16996-P-A, Rev. 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," (Reference 1) and was used for LOCA reanalysis for Braidwood and Byron.

The proposed change to the emergency core cooling system (ECCS) LOCA EM for Braidwood and Byron does not involve any changes to fuel type, peaking factors, fuel structural analysis, or boric acid precipitation methodology.

This license amendment request (LAR) completes EGC's action to perform an ECCS LOCA reanalysis in accordance with 10 CFR 50.46(a)(3)(ii) as described in EGC Letter RS-12-037, "Response to Request for Information Regarding Thermal Conductivity Degradation and 10 CFR 50.46 Report," dated March 19, 2012 and as amended in EGC Letter RS-16-239, "Supplement to RS-12-037 and RS-12-087: Revision to Exelon Generation Company, LLC Commitment Relating to Large Break Loss of Coolant Accident (LBLOCA) Analysis with an NRC Approved Emergency Core Cooling System (ECCS) Evaluation Model that Explicitly Accounts for Thermal Conductivity Degradation (TCD)," dated December 14, 2016 and EGC Letter RS-19-091, "Supplement to RS-16-239: Revision to Exelon Generation Company, LLC Commitment Relating to Large Break Loss of Coolant Accident (LBLOCA) Analysis with an NRC Approved Emergency Core Cooling system (ECCS) Evaluation Model that Explicitly Accounts for Thermal Conductivity Degradation (TCD)," dated September 5, 2019.

2.0 DETAILED DESCRIPTION The proposed changes to TS 5.6.5 reflect the NRC approved LOCA methodology that was used for the LOCA reanalysis for Braidwood and Byron. Attachments Sa and Sb to this amendment request provide the markup pages of the existing TS to show the proposed change.

2 FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, it subsidiaries and/or affiliates in the United States of America and may be registered in other countries through the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

Page 1 of 10

Attachment 1 Evaluation of Proposed Changes 2.1 Proposed Changes The proposed TS changes are described below.

TS 5.6.5.b.6 through 5.6.5.b.8 currently states:

6. WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005.

7. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985.

8. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using NOTRUMP Code," August 1985.

Revised TS 5.6.5.b.6 through 5.6.5.b.8 will state: .

6. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

7. Not Used.

8. Not Used.

3.0 TECHNICAL EVALUATION

3.1 Compliance with the Limitations and Conditions in the Revised NRC Final Safety Evaluation (FSE) for Westinghouse WCAP-16996-P-A, Rev. 1 Attachments 3 and 4 to this amendment request show Braidwood, Units 1 and 2 and Byron, Units 1 and 2 analyses are in compliance with the Limitations and Conditions.

3.2. Changes and Corrections to the FSLOCATM EM in Westinghouse WCAP-16996-P-A, Revision 1 Westinghouse has issued letter LTR-NRC-18-30 (Reference 2) and LTR-NRC-19-6 (Reference

3) to the NRC which document several changes and corrections that have been made to the FSLOCA' EM in Westinghouse WCAP-16996-P-A, Revision 1, after the NRC issued the revised Final Safety Evaluation. Those changes and corrections that are applicable to the FSLOCA ' methods amount to minimal impacts and will be handled in accordance with 10 CFR 50.46 requirements. One additional correction involving the treatment of uncertainty in the gamma energy redistribution was identified and corrected in the analysis as outline in Section 2.3 of Attachments 3 and 4. Furthermore, any future changes or corrections to the methods or the specific Braidwood and Byron Analysis of Record being implemented here will also be in accordance with 10 CFR 50.46 requirements.

Page 2 of 10

Attachment 1 Evaluation of Proposed Changes 3.3. Compliance with 10 CFR 50.46 It must be demonstrated that there is a high level of probability that the following criteria in 10 CFR 50.46 are not exceeded:

Peak Cladding Temperature (10 CFR 50.46(b)(1))-The analysis Peak Cladding Temperature (PCT), corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200°F, the analyses with the FSLOCA '

EM confirm that 10 CFR 50.46 acceptance criterion (b)(1), i.e., that PCT not exceed 2,200°F, is satisfied.

The results are shown in Tables 8 and 9 of Attachments 3 and 4 for Braidwood and Byron Unit 1 and Unit 2, respectively.

Maximum Cladding Oxidation (10 CFR 50.46(b)(2)) - The analysis Maximum Local Oxidation (MLO) corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient corrosion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e.,

that the MLO of the cladding not exceed 17 percent of the total cladding thickness before oxidation, is satisfied.

The results are shown in Tables 8 and 9 of Attachments 3 and 4 for Braidwood and Byron Unit 1 and Unit 2, respectively.

Maximum Hydrogen Generation (10 CFR 50.46(b)(3)) - The analysis Core-Wide Oxidation (CWO) corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., that CWO not exceed 1 percent of the total hypothetical amount, is satisfied.

The results are shown in Tables 8 and 9 of Attachments 3 and 4 for Braidwood and Byron Unit 1 and Unit 2, respectively.

Coo/able Geometry (10 CFR 50.46(b)(4)) - This criterion requires that the calculated changes in core geometry are such that the core remains in a coolable geometry.

This criterion is met by demonstrating compliance with criteria (b)(1 ), (b)(2), and (b)(3), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. Tables 8 and 9 of Attachments 3 and 4 show Criteria (b)(1), (b)(2), and (b)(3) have been met for Braidwood and Byron Units 1 and 2, respectively.

Section 32.1 of the NRC approved FSLOCA' EM (WCAP-16996-P-A, Revision 1) documents that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to the combined LOCA and seismic loads was calculated to not occur for Braidwood and Byron.

The FSLOCA TM EM analyses did not invalidate the existing seismic/ LOCA analysis.

Page 3 of 10

Attachment 1 Evaluation of Proposed Changes Long-Term Cooling (1 O CFR 50.46(b)(5)) - This criterion requires that long-term core cooling be provided following the successful initial operation of the EGGS.

Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRG approved FSLOCA ' EM (WCAP-16996-P-A, Revision 1).

In summary, based on the analysis results for the small-break LOCA (SBLOCA, Region I) and large-break LOCA (LBLOCA, Region II) presented in Tables 8 and 9 of Attachments 3 and 4 for Braidwood and Byron Unit 1 and Unit 2, respectively, and the discussions above relative to the criteria in 10 CFR 50.46(b)(4) and (b)(5), it is concluded that Braidwood, Units 1 and 2 and

. Byron, Units 1 and 2 would continue to comply with the criteria in 10 CFR 50.46 upon approval of this LAR.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Section 182.a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of the TS are contained in 10 CFR 50.36, "Technical specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.

However, the rule does not specify the particular requirements to be included in a plant's TS.

Under 10 CFR 50.36(c)(2)(ii), a limiting condition for operation must be included in TS for any item meeting one or more of the following four criteria:

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

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

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

The proposed change does not impact the TS safety limits, limiting safety system settings, and limiting control settings; LCOs; surveillance requirements; or design features.

Page 4 of 10

Attachment 1 Evaluation of Proposed Changes The proposed replacement for NRC approved LOCA methodology will be included in the Administrative Controls section of the Braidwood and Byron TS and would be used to determine a core operating limit. The use of the proposed NRC approved LOCA methodology will continue to ensure that the plant is operated in a safe manner. Therefore, the proposed change is consistent with the Administrative Controls requirement of 10 CFR 50.36(c)(5).

10 CFR 50.46 includes requirements and acceptance criteria pertaining to the evaluation of post accident ECCS performance. This regulation includes the requirement that" ... uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated-results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria ... there is a high level of probability that the criteria would not be exceeded."

The proposed change requests NRC approval to use the FULL SPECTRUM' LOCA (FSLOCA ') methodology described in WCAP-16996-P-A, Revision 1 for the performance of full spectrum LOCA analyses, including treatment of uncertainties in the inputs used for the analysis. No change is proposed to the analysis acceptance criteria specified in 10 CFR 50.46.

The NRC has reviewed WCAP-16996-P-A, Revision 1 and found it acceptable for referencing in licensing applications for Westinghouse designed four loop Pressurized Water Reactors.

WCAP-16996-P-A, Revision 1 is applicable to Braidwood Station and Byron Station, and the plant specific application of the FSLOCA ' methodology to the LOCA analyses have been performed in accordance with the conditions and limitations of the topical report and the associated NRC safety evaluation. The plant specific analyses demonstrate that the requirements of 10 CFR 50.46 will continue to be met, thus ensuring continued safe plant operation.

NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific, Parameter Limits from Technical Specifications," dated October 4, 1988 (Reference 4), provides that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle specific COLR that is required to be submitted to the NRC every operating cycle or each time it is revised.

Consistent with the guidance in NRC GL 88-16, Braidwood and Byron TS 5.6.5, "Core Operating Limits Report (COLR)," requires the following:

An NRC approved methodology is to be used to determine the core operating limits listed in TS 5.6.5.a;

The specific NRC approved methodologies used to determine the core operating limits are to be listed in TS 5.6.5.b; and

The COLR, including any midcycle revisions or supplements, is to be provided upon issuance for each reload cycle to the NRC in accordance with TS 5.6.5.d.

The COLR is defined in Section 1.1 of the TS and the reporting requirements in TS 5.6.5 require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised.

The GL also required that the TS include a list of references of the NRC approved Page 5 of 10

Attachment 1 Evaluation of Proposed Changes methodologies that are used to determine the cycle specific core operating limits. TS.5.6.5.b identifies the NRC approved analytical methodologies that are used to determine the core operating limits for Braidwood, Units 1 and 2 and Byron, Units 1 and 2. Upon approval of the proposed change, the guidance in the GL continues to be met since the proposed change will continue to specify the NRC approved methodologies used to determine the core operating limits.

Therefore, the proposed change to replace the previous NRC approved LOCA methodologies with the NRC approved LOCA methodology in WCAP-16996-P-A, Rev. 1, which was used for the Braidwood and Byron LOCA reanalysis, satisfies NRC GL 88-16.

4.2 Precedent The NRC has approved the following similar license amendment request to revise Core Operating Limits Report for Full Spectrum Loss-of Coolant Accident Methodology:

Letter from B.K. Singal (NRC) to J.M. Welsch (PG&E), "Diablo Canyon Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 234 and 236 to Revise Technical Specification 5.6.5b, "Core Operating Limits Report (COLR)," for Full Spectrum Loss-of-Coolant Accident Methodology (EPID L-2018-LLA-0730)," dated January 9, 2020 (ML19316A109). (Reference 5)

The following are three other similar amendment requests currently being reviewed by the NRC:

Letter from J.T. Polickoski (Tennessee Valley Authority) to US NRC, "Application to Implement the FULL SPECTRUM LOCA TM (FSLOCA ') Methodology for Loss-of-Coolant Accident (LOCA)

Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04)," dated January 17, 2020 (ML20017A338). (Reference 6)

Letter from M.D. Sartain (Dorninion Energy Virginia) to US NRC, "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (FSLOCA)," dated October 30, 2019 (ML193090197). (Reference 7)

Letter from M.D. Sartain (Dominion Energy Virginia) to US NRC, "Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident (LBLOCA)," dated October 30, 2019 (ML193090196). (Reference 8) 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC, (EGC) requests amendments to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, (Braidwood) and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron). This amendment request proposes to revise Technical Specifications (TS) 5.6.5, "Core Operating Limits Report (COLR)." The proposed change revises TS 5.6.5 to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies with a single, newer NRC approved LOCA methodology, the FULL Page 6 of 10

Attachment 1 Evaluation of Proposed Changes SPECTRUM' LOCA Evaluation Model (FSLOCA' EM), that is contained in WCAP-16996-P-A, Rev. 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," and was used for LOCA reanalysis for Braidwood and Byron. The NRC approved WCAP-16996-P-A, Revision 1 in a safety evaluation dated October 19, 2016.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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

Response: No.

The proposed change revises TS 5.6.5.b to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies listed in TS 5.6.5.b with another NRC approved methodology contained in WCAP-16996-P-A, Rev. 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)."

The proposed changes to the TS 5.6.5.b core operating limits methodologies, consists of replacing three current LOCA methodologies with a newer, single NRC approved methodology (the FSLOCA TM EM). The NRC review of the FSLOCA TM EM concluded that the analytical methods are acceptable as a replacement for the current LOCA analytical methods listed in TS 5.6.5.b.

The proposed change does not affect the design or function of any plant structures, systems, and components (SSCs). Thus, the proposed change does not affect plant operation, design features, or the capability of any SSC to perform its safety function. In addition, the proposed change does not affect any previously evaluated accidents in the UFSAR, or any SSCs, operating procedures, and administrative controls that have the function of preventing or mitigating any accident previously evaluated in the UFSAR. Thus, the proposed use of the FSLOCA TM EM will continue to assure that the plant operates in the same safe manner as before and will not involve an increase in the probability of an accident.

Page 7 of 10

Attachment 1 Evaluation of Proposed Changes The analyses results determined by use of the proposed new methodology will not increase the reactor power level or the core fission product inventory and will not change any transport assumptions or the shutdown margin requirements of the Braidwood and Byron TS. As such, Braidwood and Byron will continue to operate within the power distribution limits and shutdown margins required by the TS and within the assumptions of the safety analyses described in the UFSAR. As such, the proposed changes do not involve a significant increase in the consequences of an accident.

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 accident from any accident previously evaluated?

Response: No.

The proposed change revises TS 5.6.5.b to replace the current NRC approved Loss-of-Coolant Accident (LOCA) methodologies listed in TS 5.6.5.b with a single, newer NRC approved methodology contained in WCAP-16996-P-A, Rev. 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)." The NRC review of the FSLOCATM EM concluded that the analytical methods are acceptable as a replacement for the current LOCA analytical methods listed in TS 5.6.5.b.

The proposed change provides revised analytical methods and does not change any system functions or maintenance activities. The change does not involve physical alteration of the plant; that is, no new or different type of equipment will be installed. The change does not impact the ability of any SSC to perform its safety function consistent with the assumptions of the safety analyses and continues to assure the plant is operated within safe limits. As such, the proposed change does not create new failure modes or mechanisms that are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed change. Therefore, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The NRC has reviewed and approved the new methodology for the intended use in lieu of the current methodologies; thus, the margin of safety is not reduced due to this change.

Page 8 of 10

Attachment 1 Evaluation of Proposed Changes Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, EGC concludes that the proposed change does not involve a 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.

4.4 Conclusions 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.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated the proposed amendment and has determined that 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

1. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

2. LTR-NRC-18-30, "10 CFR 50.46 Annual Notification and Reporting for 2017," July 2018.

3. LTR-NRC-19-6, "10 CFR 50.46 Annual Notification and Reporting for 2018," February 7, 2019.

4. NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," October 1988.

5. Letter from B.K. Singal (NRC) to J.M. Welsch (PG&E), "Diablo Canyon Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 234 and 236 to Revise Technical Specification 5.6.5b, "Core Operating Limits Report (COLR)," for Full Spectrum Loss-of-Coolant Accident Methodology (EPID L-2018-LLA-0730)," dated January 9, 2020 (ML19316A109).

Page 9 of 10

Attachment 1 Evaluation of Proposed Changes

6. Letter from .T. Polickoski (Tennessee Valley Authority) to US NRC, "Application to Implement the FULL SPECTRUM LOCA TM (FSLOCA ') Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04)," dated January 17, 2020 (ML20017A338).

7. Letter from M.D. Sartain (Dominion Energy Virginia) to US NRC, "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (FSLOCA)," dated October 30, 2019 (ML19309D197).

8. Letter from M.D. Sartain (Dominion Energy Virginia) to US NRC, "Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request*

Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident (LBLOCA)," dated October 30, 2019 (ML19309D196).

Page 10 of 10

Attachment 2 Affidavit of Withholding, Proprietary Information Notice, and Copyright Notice for FSLOCATM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2"

Westinghouse Non-Proprietary Class 3 CA W-19-4935 Page 1 of 3 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Camille T. Zazula, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of Westinghouse Letter CAE-19-26 / CCE-19-25, Attachment 2 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of th~ criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses.

Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 CAW-19-4935 Page 2 of3 AFFIDAVIT (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or co,mponent, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

( c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters

Westinghouse Non-Proprietary Class 3 CA W-19-4935 Page 3 of3 AFFIDAVIT refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5Xa) through (f) of this Affidavit.

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjwy that the foregoing is true and correct.

Executed on: ,l;~Stfif.U>lf}

Camille T. Zo,, ,r

' ' ;Manager Infrastructure & Facilities Licensing

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted).

The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (5)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1 ).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Attachment Sa Braidwood Station, Units 1 and 2 NRC Docket Nos. 50-456 and 50-457 Proposed Technical Specification Changes (Mark-Up)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT CCOLR) (continued)

6. WCAP 16009 PA , Revision 0, "Realistic Large Break LOCA WCAP-16996-P-A, Evaluation Methodology Using the Automated Statistical Revision 1, Treatment of Uncertainty Method (ASTRUM) ," January "Realistic LOCA 2005 . ~

Evaluation

'n'CAP 10079 P A, " NOTRU~4P , A Nodal Transient Small Break Methodology and General NebJork Code , " August 1985 . '-F4Not Used. !

Applied to the Full Spectrum of Break h'CAP 10054 P A, " 1n'estinghouse Small Break EGGS Sizes (FULL Evaluation Model using ~JOTRUHP Code , " August 1985 .

SPECTRUM LOCA 9. WCAP-10216-P-A, Revi sion 1, "Relaxation of Constant Methodology) ," Axial Offset Control - F0 Surveillance Technical November 2016. Specification," February 1994 .

10. WCAP-8745-P-A, "Design Bases for the Thermal Overpower

.::1T and Thermal Overtemperature .::1T Trip Functions,"

September 1986.

11. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraul ic Safety Analysis," October 1999.

12. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995, (Westinghouse Proprietary).

13. WCAP-12610-P-A &CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOlM," July 2006, (Westinghouse Proprietary).

c. The core operating limits shall be determined such that all applicable li mits (e.g . , fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems CECCS) l imits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

BRAIDWOOD - UNITS 1 &2 5.6 - 4 Amendment -9G

Accumulators B 3. 5.1 BASES APPLICABLE SAFETY ANALYSES (continued) large break must be considered , and The largest break ,........, In performing the LOCA calculat ons, conservative assumptions are made concernin he availability of ECCS area considered for flow. In the early stages of a LOCA, with or without a loss a large break LOCA of offsite power , the accumulators provide the sole source is a double ended of makeup water to the RCS . The assumption of loss of guillotine break in offsite power is required by regulations and conservatively the RCS cold leg .

imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence . In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

,v The liffiiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant puffip . During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

(for loss of offsite power assumption) ,

safety injection signal generation , generation . During t his time, the accumulators are analyzed as providing the sole squrce of emergency core cooling No operator action is assumed during the bl oHdmm stage o a ,.....fo_r_a_l_a-rg-e--.

jin the modeling I large break LOCA. ~ break LOCA is assumed to inject The ~viorst case sma 11 break LOCA analyses al so assume a time into the reactor e ay e ore pumpe reaches the core . For the 1arger range of sffiall breaks, the rate of blowdown is such that the coolantsystem *ncrease in fuel clad temperature is terminated solely by

. . the accumulators, with pumped flow then providing continued 1ntermed1ate cooling. As break size decreases, the accuffiul ators and into small break f-21 centrifugal charging puffips both play a part in terminating the rise in clad temperature . As break size continues to LOCA, the decrease, the role of the accumulators continues to decrease accumulators, until they are not required and the centrifugal charging~

centrifugal pumps become solely responsible for terminating the charging and SI temperature increase . and SI pumps all BYRON - UNITS 1 &2 83.5 .1-3 Revision B

Accumulators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 4) will be met following a LOCA:

During a small break Loss Of Coolant Accident (LOCA) maximum fuel element cladding temperature is s 2200°F and during a large break LOCA there must be a high There is a high level of probability that maximum fuel element cladding temperature is s 2200°F ;

level of probability that the maximum Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation; Maximum hydrogen generation from a zirconium water There is a high reaction i s s 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding level of probability cylinders surrounding the fuel, excluding the cladding that the maximum surrounding the plenum volume, were to react; and

d. Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase large break LOCA and of -LGGA, they do not contribute to the long term cooling re irements of 10 CFR 50.46.

the recovery phase of a small break LOCA For the small break LOCA analyses , a nominal contained r:::::il A accumulator vJater volume is used . For the large break LOCA

~ analyses, a contained accumulator water volume range of

.-p-e-ra_p_p_r-ov_e_d~~-, 920 ft 3 - 980 ft 3 is use The contained water volume is t e same as e e 1vera le volume for the accumulators, methods (Ref.5) si nee the accumulators are emptied, once discharged . fe-r-small breaks , the peak cl ad temperature is not sensitive to the accumulator vJater volume . For large breaks , there are tHo competing effects regarding accumulator Hater volume:

the amount of Hater available for injection versus the injection rate . A higher ~.*ater volume results in a larger total injection but at a slmJer injection rate . Conversely ,

a lmJer Hater voluFAe results in a sFAaller total injection but at a faster injection rate .

BYRON - UNITS 1 &2 B3 .5.l-4 Revision +/-8

Accumulators B 3.5.1 BASES BACKGROUND (continued)

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series.

The motor operated isolation valves are interlocked by P-11 with the pressurizer pressure measurement channels to ensure that the valves will automatically open as RCS pressure increases to above the permissive circuit P-11 setpoint .

This interlock also prevents inadvertent closure of the valves during normal operation prior to an accident. The valves will automatically open, however, as a result of an SI signal. These features ensure that the valves meet the requirements of the Institute of Electrical and Electronic Engineers CIEEE) Standard 279-1971 (Ref. 1) for "operating bypasses" and that the accumulators will be available for injection without reliance on operator action.

The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent ILOCA transient h with the LOCA assumpti on that the entire contents of one

. , ~ accumulator will be lost vi a the RCS pipe break during the bl m~dmm phase of the LOCA .

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Refs. 2 and 3).

These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

BYRON - UNITS 1 &2 83.5.1-2 Revision B

F0 CZ)

B 3.2.1 BASES BACKGROUND (continued)

Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR Conly when PDMS is inoperable), and control rod insertion.

APPLICABLE This LCO precludes core power distributions that violate SAFETY ANALYSES the following fuel design criteria:

a. Duri ng a small break Loss Of Coolant Accident (LOCA) the peak cladding temperature must not exceed 2200°F and during a large break LOCA there must be a high 10 CFR 50.46 level of probability that the peak cladding acceptance criteria temperature does not exceed 2200°F (Ref . l);

must be met b. During a loss of forced reactor coolant flow accident, the re must be at least 95% probabi li ty at t he 95%

confidence level (the 95/95 Departure from Nucleate Boiling CDNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;

c. During an ejected rod accident, the prompt energy deposition to the fuel must not exceed 200 cal/gm (Ref. 2); and

d. The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on Fq(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains val id. Other criteria must also be met (e .g. , maximum cladding oxidation , maximum hydrogen generation , coolable geometry , and long term cooling) . Hmo*ever , the peak cladding temperature is typically most limiting .

FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to Ci .e., lower than) the F0 (Z) limit assumed in safety analyses for other postulated accidents.

Therefore, this LCO provides conservative limits for other postulated accidents.

F0(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BRAIDWOOD - UNITS 1 &2 B3.2.1-2 Revision -l-ct

F0(Z)

B 3.2.1 BASES LCO The Heat Flux Hot Channel Factor, F0 (Z), shall be limited by the following relationships:

FRTP Fa (Z) ~ - 0p- K(Z) for P > 0.5 FRTP Fa (Z) ~ _a_ K(Z) for P ~ 0.5

0.5 where

F~TP is the F0 (Z) limit at RTP provided in the COLR, K(Z) is the normalized Fg_(Z) as a function of core height provided in the CuLR, and p = THERMA L POWER RTP For th i s facility, the actual values of F'rP and K(Z) are given in the COLR; however, F'rP is normally a number on the order ~ . and K(Z) is a function that looks like the 2 .60 o rovided i n Figure B 3.2.1-1.

F0 (Z) is approximated by FgCZ) and F;(Z) . Thus, both Fg(Z) and F; (Z) must meet the preceding l i mi ts on FaC Z) .

When PDMS is inoperable, an FgCZ) evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value CF~CZ)) of FaC Z). Then, Fg (Z) = F~ (Z)

Cl. 0815) where 1.0815 is a factor that accounts for fuel manufacturing tolerances and flux map measurement uncertainty.

Fg CZ) is an excellent approximation for Fa( Z) when the reactor is at the steady state power at which the incore flux map was taken.

BRAIDWOOD - UNITS 1 &2 B 3.2.1-3 Revision tt

F0 CZ)

B 3.2.1 BASES LCO (continued)

When PDMS is OPERABLE, FaCZ) is determined continuously.

Then, F5(Z) = F~(Z)

UFO where U~ is a factor that accounts for measurement uncertainty (Ref. 4) and engineering uncertainty defined in the COLR.

The expression for F~(Z)is :

F~(Z) = F5(Z)

W(Z) where WCZ) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR. When PDMS is inoperable, the FgCZ) is calculated at equilibrium conditions .

ensure that the This LCO requires operation within the bounds assumed in the 10 CFR 50.46 safety anal yses . Calculations are performed in the core acceptance criteria design process to confirm that the core can be controlled in are met such a manner during operation that it can stay within the LOCA F0 (Z) limits. If Fg(Z) cannot be maintained within the LCO limits , reduction of the core power is required .

Violating the LCO limits for F0(Z) may produce unacceptable consequences if a design basis event occurs while F0 (Z) is outside its specified limits.

APPLICAB ILITY The F0(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses. Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power .

BRAIDWOOD - UNITS 1 &2 B 3.2.1 - 4 Revision B

F0 (Z)

!This figure is replaced with the figure on the next page. B 3.2 .1 9 ,--- ** .

1

-~ I I :~ ~N~~!~~~~~~INTH,SAREA~

1

-1 I

1 I I ;' .Ii

--1----1

: ' ' i 1.0~ 1--+--~1 - -(_6._0.r\- 1-..:

.0 )

I i I ( 12( 0 .925)

I I I I

0. B ____ __..... . . . **- l l

. I I I Na 6

~*I L_ -. -- l----

I

-* - ---. . r-.. -- -- -* --~ ---- - --*

1 I

I I i I 0.41--- i----:-+/- .-~-:-

J--~--, _l .

I I IL[~~~~ifS~ b~~Y.

I DO NOT USE FOR

..... r*-I

I.

I 0 .2 -

I OPERATION REFER TO COLR FOR .

ACTIJAL VALUES r

I 1

I I I I

o~ - 2 ***- --_4 - --t-"---~---*-* - fo- - *-/2 CORE HEIGHT (F'T.)

Figure B 3.2.1-1 (page 1 of 1)

K(Z) - Normalized F0(Z) as a function of Core Height BRAI DWOOD - UN ITS 1 &2 B 3.2.1 - 13 Revi sion 2-

1.2 DO NOT OPERA TE IN THIS AREA (0.0, 1.0) (6.0, 1.0) (12 .0, 1.0) 1.0 0.8

~0.6 0.4 THIS FIGURE FOR ILLUSTRATION ONLY.

DO NOT USE FOR 0.2 OPERATION.

REFER TO COLR FOR ACTUAL VALUES .

0.0 0 2 4 6 8 10 12 CORE HEIGHT (FT.)

Figure B 3 . 2 . 1-1 (page 1 of 1)

K(Z) - Normalized F0 (Z) as a function of Core Height

BASES BACKGROUND (continued)

The COLR provides peaking factor limits that ensure that the design criterion for the Departure from Nucleate Boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. All DNB limited transient events are assumed to begin with an F1H value that satisfies the LCO requirements.

Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.

APPLICABLE Limits on F1H preclude core power distributions that exceed SAFETY ANALYSES the following fuel design limits:

a. There must be at least 95% probability at the 95%

confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition ;

b. During a small break Loss Of Coolant Accident (LOCA)

Peale Cladding Temperature (PCT) must not exceed 2200°F the 10 CFR 50.46 and during a large break LOCA there must be a high level of probabi li ty that PCT does not exceed 2200°F; acceptance criteria must be met c. During an ejected rod accident, the prompt energy (Ref.3) deposition to the fuel must not exceed 200 cal/gm (Ref . 1) ; and

d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when cont rol rods must be capable of shutting down the reactor with a minimum required Shutdown Margin with the highest worth control rod stuck fully withdrawn.

For transients that may be DNB limited, F1H is a significant core parameter. The limits on F1H ensure that the DNB design crite ri on is met for normal operation, operational transients , and any transients arising from events of moderate frequency. Refer to the Bases for LCO 3 .4 . 1, "RCS Pressure, Temperature, and Fl ow DNB Limits," for a discussion of the applicable Departure from Nucleate Boiling Ratio (DNBR) limits.

BRAIDWOOD - UNITS 1 &2 B 3.2.2 - 2 Revision BASES APPLICABLE SAFETY ANALYSES (continued)

The allowable F1H limit increases with decreasing power level. This functional ity in F1H is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of F1H in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial F1H as a function of power level defined by the COLR limit equation .

The Nuclear Enthalpy Rise Hot The LOCA safety analysis indirectly FRodels -F,!. as an input If\ paraFReter . The Nuclear Heat Flu)( Hot Channel Factor (FG(Z))

Channel Factor and the a)(ial peaking factors are inserted directly into the (FNt,H), the Nuclear LOCA safety analyses that verify the acceptability of the Heat Flus Hot resulting peak cladding teFRperature (Ref . 3) .

Channel Factor (Fa(Z)), and the axial The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the peaking factors are U safety and accident analyses remain valid. The following supported by the LCOs ensure this: LCD 3.1.6, "Control Bank Insertion LOCA safety Limits," LCD 3.2.1, "Heat Flux Hot Channel Factor (F0 (Z)),"

analyses that verify LCD 3.2.2, LCD 3.2.3, LCD 3.2.4, and LCD 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)."

compliance with the 10 CFR 50.46 F~ and ~(Z) are measured periodically using the movable acceptance criteria incore detector system when PDMS is inoperable.

(Ref. 3). Measurements are generally taken with the core at, or near, steady state conditi ons. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR , and Control Bank Insertion Limits . When PDMS is OPERABLE, F1H and F0(Z) are determined continuously . Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on DNBR and Control Bank Insertion Limits.

F1H satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) .

BRAIDWOOD - UNITS 1 &2 B 3.2.2 - 3 Revision ft

F~H B 3.2.2 BASES LCO F~H shall be maintained within the limits of the relationship provided in the COLR.

The F~H limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB.

The limiting value of F1H, described by the equation contained in the COLR, is the design radial peaking factor used in the plant safety analyses.

The power multiplication factor in this equation provides margin for higher radi al peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of F1H is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER .

APPLICABILITY The F1H limits must be maintained in MODE 1 to prevent core power distributions from exceeding the fuel design limits for DNBR a -Pf+. Appl icability in other modes is not the 10 CFR 50.46 requir e ause there is either insufficient stored energy acceptance criteria i e fue l or i nsufficient energy being transferred to the reactor coolant t o require a limit on the distribution of (Ref. 3) core power . Specifically , the design bases events that are sensitive to F1H in other modes (MODES 2 through 5) have significant margin to DNB, and therefore, there is no need to restrict F1H in these modes .

ACTIONS A.1. A.2. A.3. and A.4 With F1H exceeding its 1i mit , Condi ti on A is entered . F1H may be restored to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

through, for example, real igning any misaligned rods or reducing power enough to bring F1H within its power dependent limit . If the value of F1H is not restored to within its specified limit, THERMAL POWER must be reduced to

< 50% RTP in accordance with Required Action A.1. When the F~H limit is exceeded , the DNBR limit is not likely violated in steady state operation , because events that could significantly perturb the F1H value (e.g., static control rod misalignment) are considered in the safety analyses .

BRAIDWOOD - UNITS 1 &2 B 3.2.2 - 4 Revision tt

QPTR B 3 . 2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodical ly during power operation.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. When Power Distribution Monitoring System (PDMS) is OPERABLE, Peak Linear Heat Rate and the linear power along the fuel rod with the highest integrated power are measured continuously .

APPLICABLE Limits on QPTR preclude core power distributions that SAFETY ANALYSES violate the following fuel design criteria:

a. During a small break Loss Of Coolant Accident (LOCA) the Peak Cladding Temperature (PCT) must not exceed 10 CFR 50.46 2200°F and during a large break LOCA there must be a high l evel of probability that t he PCT does not e)(ceed acceptance criteria 2200°F ( Ref. 1) ;

must be met

b. During a loss of forced rea ctor coolant flow accident ,

there must be at least 95% probability at the 95%

confidence level (the 95/95 Departure from Nucleate Boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;

c. During an ejected rod accident, the prompt energy deposition to the fuel must not exceed 200 cal/gm (Ref. 2) ; and BRAIDWOOD - UNITS 1 &2 B 3.2.4 - 1 Revision ESFAS Instrumentation B 3.3.2 BASES APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY ANALYSES, more ESFAS Functions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident . An ESFAS APPLICABILITY Function may be the primary actuation signal for more than one type of accident . An ESFAS Function may also be a secondary, or backup, actuation signal for one or more other accidents . For example, Pressurizer Pressure-Low is a primary actuation signal for smal l Loss Of Coolant Accidents (LOCAs) and a backup actuation signal for Steam Line Breaks (SLBs) outside containment . Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit.

These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions may also serve as backups to Functions that were credited in the accident analysis (Ref . 3).

The LCO requires all instrumentation performing an ESFAS Function to be OPERABLE when the unit status is within the Applicability . A channel is OPERABLE with a trip setpoint outside its calibration tolerance band provided the trip setpoint "as found" value does not exceed its associated A11 owab le Value and provided the trip setpoi nt "as left" value is adjusted to a val ue within the calibration tolerance band of the Nominal Trip Setpoint . A trip setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions .

Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions .

The LCO generally requires OPERABILITY of three or four channels in each instrumentation Function and two channels in each logic and manual initiation Function . The two-out -of-three and the two-out-of-four configurations allow one channel to be tripped during maintenance or testing without causing an ESFAS initiation . Two logic or manual initiation channels are required to ensure no single random failure disables the ESFAS .

BRAIDWOOD - UNITS 1 &2 B 3.3.2 - 6 Revision re

ESFAS Instrumentation B 3.3 .2 BASES APPLICABLE SAFETY ANALYSES, LCD, and APPLICABILITY (continued)

The required channels of ESFAS instrumentation provide unit protection in the event of any of the analyzed accidents.

ESFAS protection functions are as follows :

1. Safety In.iecti on Safety Injection (SI) provides two primary functions :

1. Primary side water addition to ensure maintenance or recovery of reactor vessel water level (coverage of the active fuel for heat removal ,

clad integrity, and for limiting peak clad compliance with temperature to < 2200°F) ; and the 10 CFR 50.46

2. Borati on t o ensure recovery and maintenance of acceptance criteria SOM.

(Ref. 17)

These functions are necessary t o mi tigate the effects of High Energy Line Breaks (HELBs) both inside and outside of containment . The SI signal is also used to initiate other Functions such as :

Phase A Isolation ;

Containment Purge Isolation;

Reactor Trip ;

Turbine Trip;

Feedwater Isolation ;

Start of Auxiliary Feedwater (AF) pumps;

Control room ventilation isolation ; and

Enabling automatic switchover of Emergency Core Cooling Systems (ECCS) pump suction to containment sump.

BRAIDWOOD - UNITS 1 &2 B 3.3.2 - 7 Revision .ft

ESFAS Instrumentation B 3.3.2 BASES REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 7.

3. UFSAR, Chapter 15.

4. IEEE-279-1971.

5. Technical Requirements Manual.

6. WCAP-12523, "Bases Document for Westinghouse Setpoint Methodology for Protection Systems, Zion/Byron/Braidwood Units" October 1990.

7. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.

8. WCAP-13632 Revision 2, "El imi nation of Pressure Sensor Response Time Testing Requirements," August 1995.

9. UFSAR, Section 7.3.

10. WCAP-12583, "Westinghouse Setpoint Methodology For Protection Systems, Byron/Braidwood Stations," May 1990.

11. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

12. Not used.

13. Not used .

14. Not used.

15. WCAP-14333-P-A, Revision l, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," October 1998.

16. WCAP-15376-P-A, Revision l, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2000.

117. 10 CFR 50.46 . I BRAIDWOOD - UNITS 1 &2 B 3.3.2 - 58 Revision +/-8 Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS CECCS)

B 3.5 .1 Accumulators large break BASES BACKGROUND The functions of the ECCS accumulators are to suppl ater to the reactor vessel during the blowdown phase of a Loss Of Coolant Accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphe ~

In the refill phase of a 4:ocA, ~ ately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup . The balance of accumulator inventory is then available t elp fill voids in the lower plenum and reactor vessel down omer so as to establish a recovery level at the bottom oft e core a ongoing reflood of the core with the addition f Sa njection (SI) water.

initiate The accumulators are pressure vessels partially filled w th borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tan k pressure is sufficient to discharge the accumulator contents to the RCS if RCS pressure decreases below the accumulator pressure.

Initial accumulator inventory which is injected into the reactor vessel is lost out the break.

BRAIDWOOD - UNITS 1 &2 B 3.5.1 - 1 Revision {j.

Accumulators B 3.5.1 BASES BACKGROUND (continued)

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series .

The motor operated isol ation valves are interlocked by P-11 with the pressurizer pressure measurement channels to ensure that the valves wil l automatically open as RCS pressure increases to above the permissive circuit P-11 setpoint .

This interlock also prevents inadvertent closure of the valves during normal operation prior to an accident . The valves will automatically open, however, as a result of an SI signal. These features ensure that the valves meet the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) for "operating bypasses" and that the accumulators will be available for injection without reliance on operator action.

The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent

!LOCA transient h with the LOCA assumption that the entire contents of one

~ accumulator will be lost vi a the RCS pipe break during the bl m.dmm phase of the LOCA .

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Refs. 2 and 3) .

These are the Design Basis Accidents CDBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

BRAIDWOOD - UNITS 1 &2 B3.5.1-2 Revision {}

Accumulators B 3. 5.1 BASES

- - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - 1 must be APPLICABLE SAFETY ANALYSES (continued) large break considered, In performing the LOCA calculati ns, conservative ~a_n_d~ _ ___.

assumptions are made concernin he availability of ECCS The largest break -- fl ow. In the early stages of a LOCA, with or without a l os area considered for of offsite power, the accumulators provide the sole source a large break LOCA of makeup water to the RCS . The assumption of loss of is a double ended offsi te power is required b:Y regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow guillotine break in until the emergency diesel generators start, come to rated the RCS cold leg . speed, and go through their timed loading sequence . In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break .

,11 The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump . During this event, the accumulators discharge to the RCS as soon as (for loss of offsite RCS ressure decreases to below accumulator pressure .

power assumption) ,

As a conservative estimate , no redit is taken for ECCS pump safety injection flow until an effective delay ha elapsed. This delay signal generation , accoun s o the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set ~Jith an additional 2 seconds to account for SI signal generation . During this time , t he accumulators are ana lyzed as providing the sole source of emergency core cooling.

operator action is assume during the bl mi'clmm stage of a in the modeling la rge break LOCA . for a large

. break is assumed to inject The vi*orst case small br ak LOCA analyses al so assume a time LOCA e ay e ore pumpe reaches the core . For the larger i=.::c....=....:......._ _,

into the reactor range of sFRall breaks , the rate of blowdown is such that the coolant system increase in fue l clad temperature is terminated solely by intermediate the accumulators , with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging puFRps both play a part in terminating into small break the rise in clad temperature. As break size continues to LOCA, the decrease , the rol e of the accumulators continues to decrease accumulators, until they are not required and the centrifugal charging ~

centrifugal pumps become solely responsible for terminating the temperature increase. and SI charging and SI pumps all BRAIDWOOD - UNITS 1 &2 83 .5.1-3 Revision B-

Accumulators B 3. 5.1 BASES APPLICABLE SAFETY ANALYSES (continued)

This LCO hel ps to ensure that the followi ng acceptance criteria established for the ECCS by 10 CFR 50 .46 (Ref . 4) wi l l be met following a LOCA :

~ a. . During a small break Loss Of Cool ant Accident ( LOCA)

~ ma> 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

This LCO is only applicable at pressures> 1000 psig . At pressures~ 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that the 10 CFR 50.46 peal< clad temperature remains belmJ the 10 CFR 50 .46 (Ref . 4) li mit of 2200°F.

(Ref. 4) acceptance criteria are met BRAIDWOOD - UNITS 1 &2 B3.5.1-6 Revision Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.1.5 Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 1% volume increase (nominally 70 gallons or 10% of indicated level) will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration of the accumulator after a 1% volume increase (10% indicated level increase) if the added water inventory is from the Refueling Water Storage Tank (RWST) and the boron concentration of the RWST is~ 2200 ppm and~ 2400 ppm.

With the water contained in the RWST within the boron concentration requirements of the accumulators, any added inventory would not cause the accumulator's boron concentration to exceed the limits of this LCO.

With the only indication available to the operators in the control room being level indication in percent, a required accumulator volume increase of 1% or an increase of 10% of indicated level would require the accumulator to be sampled to verify the accumulator boron concentration is within the limits. The small break LOCA analysis assumes a nominal water volume of 7106 gal lons based on 3the TS minimum and

~ B'f- 6995 gallons (935 ft , 31%of indicated

~ ~ 2 ' 1 7 gallons (965 ft 3 , 63%of indicated level) .

These volumes are also indicated in the specific tank curves for the SI accumulators. The large brealc LOCA analysis

~~~-~7 assumes a water vol ume range of 6882 gallons (920 ft 3 , 15%

r of indicated level) to 7331 gallons (980 ft 3 , 79% of small break and indicated level) which bounds the TS limits. The 10%

large break LOCA indicated level increase is considered a conservative analyses assume indicati on for a 70 gallon increase in the accumulator volume requiring an increase in the sampling requirement to verify accumulator boron concentration remains within the specified limits.

BRAIDWOOD - UNITS 1 &2 B 3.5.1 - 9 Revision 8&

Accumulators B 3.5.1 BASES SURVEILLANCE REQU IREMENTS (continued)

SR 3.5.1.6 Verification that power is removed from each accumulator isolation valve operator ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA.

The power to the accumulator motor operated isolation valves is removed by opening t he motor control center breaker and tagging it out administratively. The Surveillance Frequency is controlled under the Su rveillance Frequency Control Program.

REFERENCES 1. IEEE Standard 279- 1971.

2. UFSAR, Chapter 15.
3. UFSAR, Chapter 6.
4. 10 CFR 50.46.
5. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016 .
BRAIDWOOD - UNITS 1 &2 B 3.5.1 -10 Revision 85-
ECCS-Operating B 3.5.2 BASES APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria for the ECCS, established by 10 CFR 50.46 (Ref. 2),
will be met following a LOCA:
~ a. During a small break LOCA ma>(imum fuel element
~ ~ cladding temperature is~ 2200°F and during a large break LOCA there must be a high level of probability t hat maximum fue l element cladding temperature is There is a high ::; 2200°F ;
level of probability that the maximum Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation; Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount that There is a high would be generated if all of the metal in the cladding level of probability cylinders surrounding the fuel, excluding the cladding that the maximum surrounding the plenum volume, were to react;
d. Core is maintained in a coolable geometry; and
e. Adequate long term core cooling capability is maintained .
The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment temperature limits are met.
Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Ref. 3). This event establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The centrifugal charging pumps and SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The SGTR and MSLB events also credit the centrifugal charging pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions : ~
a. For the large break LOCA event , the ASTRUM methodology examines LOOP and no-LOOP cases with a single fai l ure disabling one train of SI pumps. No fail ure of containment heat removal system is modeled; and BRAIDWOOD - UNITS 1 &2 B 3.5 .2 - 4 Revision RWST B 3.5 .4 BASES APPLICABLE SAFETY ANALYSES (continued) a range of RWST ~ In the ECCS analysis, the containment spray temperature is temperatures of assumed to be equal to the RWST l o~Jer temperature limit of 32°F to 120°F is
~ - If the lower temperature limit is violated, the containment spray further reduces containment pressure. The used for the reduced containment pressure lowers the quality of steam containment spray exiting the break thus decreasing the rate which the steam temperature per is vented to the containment atmosphere . The decreased rate approved methods of steam vented to the containment atmosphere results in a (Ref.4) .
corresponding decrease in the rate the Reactor Coolant System pressure drops and the rate ECCS fluid is injected in the core thereby causing a rise in peak clad temperature.
ii1- upper temperature limit of 100°F is used in the small break LOCA analysis and containment OPERABILITY analysis.
Exceeding this temperature will result in a higher peak cl ad temperature , because there is less heat transfer from the core to the injected Hater for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.
The limits on RWST level and boron concentration also ensure that the post-LOCA sump pH wi l l be between 8.0 and 11.0.
The minimum and maximum pH values are verified for each fuel cycle using conservative maximum and minimum RWST volumes and the maximum and minimum allowed RWST boron concentrations. The LOCA offsite dose analysis assumes a conservatively low sump pH for the re-evolution of iodi ne from the sump. Ensuring that the minimum sump pH is at least 8.0 protects mechanical components and equipment inside containment from the effects of chloride induced stress corrosion cracking . Ensuring that the maximum sump pH is no greater than 11 .0 limits the production of hydrogen due to the corrosion of aluminum and zinc inside containment . Finally , the limits on RWST boron concentration also ensure that the containment spray pH is acceptable . The calculation of the iodine removal effectiveness of the containment spray assumes a conservatively low containment spray pH .
The RWST satisfies Criterion 3 of 10 CFR 50 .36(c)(2)(ii) .
~ - - - - - - - T h e small break and large break LOCA analyses use a range of RWST temperatures of 32°F to 120°F per approved methods (Ref. 4) .
Exceeding the maximum temperature of this range could result in a higher peak cladding temperature, because there is less heat transfer from the core to the injected water. The containment BRAIDWOOD - UNITS 1 &2 B 3.5.4 - 5 Revision .g.
RWST B 3.5.4 BASES REFERENCES 1. WCAP-13964, Revision 2, "Conmonwealth Edison Company, Byron/Braidwood Units 1 &2, Increased Steam Generator Tube Plugging/Reduced Thermal Design Flow/Positive Moderator Temperatu re Coefficient Analysis Program, Engineering/Licensi ng Report," September 1994.
2. UFSAR, Chapter 15.
3. UFSAR, Section 6.2.1.
4. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) ," November 2016.
BRAIDWOOD - UNITS 1 &2 B 3.5.4 - 9 Revision B
Containment Pressure B 3.6.4 BASES APPLICABLE SAFETY ANALYSES (continued)
The initial pressure condition used in the containment analysis was 1.0 psig . This resulted in a maximum peak pressure from a LOCA of 42.8 psig for Unit 1 and 38.4 psig for Unit 2. The containment analysis (Ref. 1) shows that the maximum peak calculated containment pressure, Pa, results from the limiting LOCA . The maximum containment pressure resulting from the worst case LOCA does not exceed the containment design pressure, 50 psig.
The containment was also evaluated for an external pressure load equivalent to -3.5 psig (Ref. 2). The inadvertent actuation of the Containment Spray System was analyzed to determine the resulting reduction in containment pressure.
The initial pressure condition used in this analysis was 0.0 psig. This resulted in a minimum pressure inside containment of -3.48 psig, which is less than the design load.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative . In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore , for the reflood phase ,
the containment backpressu re is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50 , Appendix K (Ref . 3) . .--F-o-r-th-e-se~ca-l~cu-1-at-io_n_s_,
Containment pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) .
WCAP-16996-P-A, Revision 1 BRAIDWOOD - UNITS 1 &2 B 3.6.4 - 2 Revision Containment Pressure B 3.6.4 BASES ACTIONS (continued)
B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR, Section 6.2.
2. Safety Evaluation Report Related to the Operation of Byron Station Units 1 and 2, Supplement 2.
3. 10 CFR ~ Appendix K.
WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) ," November 2016.
BRAIDWOOD - UNITS 1 &2 B 3.6.4 - 4 Revi sion 85-
Containment Spray and Cooling Systems B 3.6 .6 BASES APPLICABLE SAFETY ANALYSES (continued)
The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure is 42 .8 psig for Unit 1 and 38.4 psig for Unit 2 (experienced during a LOCA). The analysis shows that the peak containment temperature is 333.6°F for Unit 1 and 330.8°F for Unit 2 (experienced during an SLB). Both results meet the intent of the design basis . (See the Bases for LCD 3.6.4, "Containment Pressure," and LCD 3.6.5 for a detailed discussion . ) The analyses and evaluations assume a unit specific power level of 3672 .6 MWt, one containment spray train and one containment cooling train operating, and initial (pre-accident) containment conditions of 120°F and 1.0 psig. The analyses also assume a response time delayed initiation to provide conservative peak calculated containment pressure and temperature responses.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative . In particular , the effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure . For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize , rather than maximize, the calculated transient contai nment pressures in accordance wi 10 CFR 50 , Appendix K (Ref . 4).
WCAP-16996-P-A, Revision 1 The effect of an inadvertent containment spray actuation has been analyzed. An inadvertent spray actuation results in a -3.48 psig containment pressure and is associated with the sudden cooling effect in the interior of the leak tight containment. Additional discussion is provided in the Bases for LCD 3.6.4.
The modeled Containment Spray System actuation from the LOCA containment analysis is based on a response time associated with exceeding the containment High-3 pressure setpoint to achieving full flow through the containment spray nozzles.
The Containment Spray System total response time of 110 .2 seconds (for the l imiting case) includes Diesel Generator (DG) startup (for loss of offsite power) ,
sequencing of equipment , containment spray pump startup , and spray line filling (Ref. 5).
BRAIDWOOD - UNITS 1 &2 B 3.6 .6 - 5 Revision -94
Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.
REFERENCES 1. 10 CFR 50 , Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43 .
2. UFSAR, Section 9.4 .8.
3. UFSAR, Section 6.5.2.
4. 10 GFR 50 , Appendix K.
1'
5. UFSAR, Section 6.2.1.1.3.
6. UFSAR , Section 6.2.2.
7. UFSAR, Section 6.2.
8. ASME Code for Operation and Ma i ntenance of Nuclear Power Pl ants .
WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
BRAIDWOOD - UNITS 1 &2 B 3.6.6 - 13 Revision .W§.
AF System B 3.7.5 BASES APPLICABLE SAFETY ANALYSES (continued)
In addition, the minimum available AF flow and system characteristics are serious considerations in the analysis of a small brea k Loss Of Coolant Accident (LOCA) and loss of offsite power (Ref. 3).
The AF System design is such that it can perform its function following an FWLB between the main feedwater isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of one AF pump. The AF lines to the SGs are orificed such that sufficient flow is delivered to the non faulted SGs. Reactor trip is assumed to occur when the faulted SG reaches the low-low level setpoint. Sufficient flow would be delivered to the intact steam generators by the other AF pump.
During the loss of all AC power events, the Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AF diesel driven pump and associated controls to ensure an adequate supply to the steam generators during loss of power. Valves which can be manually controlled are provided for each AF line to control the AF flow to each steam generator during loss of all AC power events .
AF flow is also an important consideration in the Steam Generator Tube Rupture (SGTR) event (Ref. 4) . For the SGTR event , isolation of AF is an important recovery action . The flow path is normally isolated with the motor operated AF013 valves. Prior to isolation with AF013 valves, flow to the ruptured SG is limited by the control function of the AF005 valves .
The AF System satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii) .
Since the AF005 valve is a backup and diverse method to isolate flow to the faulted SG in a SGTR event, this function does not satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) . Administrative requirements for this function are addressed in the Technical Requirements Manual.
BRAIDWOOD - UNITS 1 &2 B 3.7.5 - 3 Revision ++/-8
Attachment Sb Byron Station, Units 1 and 2 NRC Docket Nos. 50-454 and 50-455 Proposed Technical Specification Changes (Mark-Up)
Containment Pressure B 3.6.4 BASES APPLICABLE SAFETY ANALYSES (continued)
The initial pressure condition used in the containment analysis was 1.0 psig . This resulted in a maximum peak pressure from a LOCA of 42 .8 psig for Unit 1 and 38.4 psig for Unit 2. The containment analysis (Ref. 1) shows that the maximum peak calculated containment pressure, Pa, results from the limiting LOCA . The maximum containment pressure resulting from the worst case LOCA does not exceed the containment design pressure, 50 psig.
The containment was also evaluated for an external pressure load equivalent to -3.5 psig (Ref. 2). The inadvertent actuation of the Containment Spray System was analyzed to determine the resulting reduction in containment pressure.
The initial pressure condition used in this analysis was 0.0 psig . This resulted in a minimum pressure inside containment of -3.48 psig, which is less than the design load .
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore , for the reflood phase ,
the containment backpressu re is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance~----..___--~
with 10 CFR 50 , Appendix K (Ref . 3) . For these calculations Containment pressure satisfies Criterion 2 of WCAP-16996-P-A, 10 CFR 50. 36 (c) (2)( i i) .
Revision 1 BYRON - UNITS 1 &2 B 3.6.4 - 2 Revision +8
Containment Pressure B 3.6.4 BASES ACTIONS (continued)
B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply . To achieve this status , the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable , based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging pl ant systems .
SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. UFSAR , Section 6.2.
2. Safety Evaluation Report Related to the Operation of Byron Station Units 1 and 2, Supplement 2.
3. 10 CFR 50 , Appendi x K.
(
WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016 .
BYRON - UNITS 1 &2 B 3.6.4 - 4 Revision +&
Accumulators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)
Both the large and the small break LOCA analyses model the pipe water volume from the accumulator to the SI accumulator discharge header downstream cold leg injection check valve (SI8948). However , an evaluation was performed neglecting the pipe water volume between the SI accumulator discharge header upstream cold leg injection check valve (SI8956) t o the SI accumulator discharge header downstream cold leg injection check valve (SI8948) to address gas accumulation.
This evaluation determined that the impact on peak clad is minimal for the tern eratu \das FRi niFRa l for both the large break and the sma re LOCA analyses. Si nee the range of the all O'vded large break LC?CA accuFRu l ator vol uFRes is relati vely sFRall and has a FRi niFRal analyses and 1s effect on peak clad teFRperature , a noFRinal water voluFRe is 20°F penalty for used in the sFRall break LOCA analysis . The sFRall break LOCA analysis assuFRes a noFRi nal *,,'ater vol uFRe of 7106 gall ens The based on the Technical Specification (TS) minimum and are maximum limi Bf. 6995 gallons (935 ft 3 , 31% of indicated
~~~~~~...._,_. level ) and 7 7 gallons (965 ft3, 63%of indicated level).
small break and Th large break LOCA analysis assuFRes a water volume range largebreakLOCA o 6882 ga llons (920 ft3, 15%of indi cated level) to 7331 analyses assume ga 11 ons (980 ft 3 , 79%of indicated level ) which bounds the TS limits .
The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particul ar interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH . The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and mini mum sump pH.
landh
~ ~ he
. small break LOCA analyses are perforFRed at the FRiniFRUFR nitrogen cover pressure , since sensitivity analyses have deFRonstrated t hat higher nitrogen cover pressure resu l ts in a coFRputed peak clad teFRperature benefit . -TAe- large break LOCA ana lyses are performed at a nitrogen cover pressure 586.3 psig to 662 .3 ran 587 psia to 692 psia . The maximum nitrogen cover psig per approved pressure limit prevents accumulator relief valve actuation ,
and ultimately preserves accumulator integrity.
methods (Ref. 5) .
BYRON - UNITS 1 &2 B 3.5 .1-5 Revision -74
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
6. ,~CAP 16009 PA , Revision 0 , " Realistic Large Break LOCA 1
WCAP-16996-P-A, Evaluation Methodology Using the Automated Statistical Revision 1, Treatment of Uncertainty Method (ASTRUM) ," January "Realistic LOCA 2005 . ~
Evaluation tJCAP 10079 P A, "NOTRUMP , A Nodal Transient Small Break 1
Methodology and General Net*,vork Code , " August 1985 . ~Not Used. !
Applied to the Full
8. 1~CAP 10054 P A, "Westinghouse Small Break EGGS 1
Spectrum of Break Evaluation Model using NOTRUMP Code ," August 1985 .
Sizes (FULL SPECTRUM LOCA 9. WCAP-10216-P-A, Revision 1, "Relaxation of Constant Methodology) ," Axial Offset Control - F0 Surveillance Technical November 2016 . Specifi cation," February 1994.
10. WCAP-8745-P-A, "Design Bases for the Thermal Overpower
~T and Thermal Overtemperature ~T Trip Functions,"
September 1986.
11. WCAP-14565-P-A, "V IPRE-01 Modeling and Qualification for Pressurized W ate r Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
12. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995, (Westinghouse Proprietary).
13. WCAP-12610-P-A &CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," July 2006, (Westinghouse Proprietary).
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits , Emergency Core Cooling Systems (ECCS) limits , nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
BYRON - UNITS 1 &2 5.6 - 4 Amendment %
Containment Spray and Cooling Systems B 3.6.6 BASES APPLICABLE SAFETY ANALYSES (continued)
The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure is 42 .8 psig for Unit 1 and 38.4 psig for Unit 2 (experienced during a LOCA). The analysis shows that the peak containment temperature is 333.6°F for Unit 1 and 330.8°F for Unit 2 (experienced during an SLB) . Both results meet the intent of the design basis. (See the Bases for LCD 3.6.4, "Containment Pressure," and LCD 3.6.5 for a detailed discussion . ) The analyses and evaluations assume a unit specific power level of 3672 .6 MWt, one containment spray train and one containment cooling train operating, and initial (pre-accident) containment conditions of 120°F and 1.0 psig. The analyses also assume a response time delayed initiation to provide conservative peak calculated containment pressure and temperature responses .
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative . In particular , the effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure . For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance wi 10 CFR 50 , /\ppendi x K (Ref . 4) .
WCAP-16996-P-A, .__~
Revision 1 The effect of an inadvertent containment spray actuation has been anal yzed . An inadvertent spray actuation results in a
-3.48 psig containment pressure and is associated with the sudden cooling effect in the interior of the leak tight containment. Additional discussion is provided in the Bases for LCO 3.6.4.
The modeled Contai nment Spray System actuation from the LOCA containment analysis is based on a response time associated with exceeding the containment High-3 pressure setpoint to achieving full flow through the containment spray nozzles .
The Containment Spray System total response time of 110 .2 seconds (for the limiting case) includes Diesel Generator COG) startup (for loss of offsite power),
sequencing of equipment , containment spray pump startup, and spray line filling (Ref . 5) .
BYRON - UNITS 1 &2 B 3.6.6 - 5 Revision g§.
Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41 ,
GDC 42, and GDC 43.
2. UFSAR, Section 9.4.8.
3. UFSAR, Section 6.5.2.
4. 10 CFR 50 , Appendix K.
5. UFSAR, Section 6.2.1.1.3 .
6. UFSAR, Section 6.2.2.
7. UFSAR, Section 6.2 .
WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) ,"
November 2016 BYRON - UNITS 1 &2 B 3.6.6 - 13 Revision -%-
AF System B 3.7.5 BASES APPLICABLE SAFETY ANALYSES (continued)
In addition, the minimum available AF flow and system characteristics are serious considerations in the analysis of a sffiall break Loss Of Coolant Accident (LOCA) and loss of offsite power (Ref . 3) .
The AF System design is such that it can perform its function following an FWLB between the main feedwater isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of one AF pump. The AF lines to the SGs are orificed such that sufficient flow is delivered to the non faulted SGs . Reactor trip is assumed to occur when the faulted SG reaches the low-low level setpoint . Sufficient flow would be delivered to the intact steam generators by the other AF pump .
During the loss of all AC power events, the Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AF diesel driven pump and associated controls to ensure an adequate supply to the steam generators during loss of power. Valves which can be manually controlled are provided for each AF line to control the AF flow to each steam generator during loss of all AC power events .
AF flow is al so an important consideration in the Steam Generator Tube Rupture (SGTR) event (Ref . 4) . For the SGTR event, isolation of AF flow is an important recovery action .
This flow path is normal ly isolated with the motor operated AF013 valves. Prior to isolation with AF013 valves , flow to the ruptured SG is limited by the control function of the AF005 valves .
The AF System satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii) .
Since the AF005 valve is a backup and diverse method to isolate flow to the fa ulted SG in a SGTR event , this function does not satisfy Criterion 3 of 10 CFR 50 .36(c)(2)(ii) . Administrative requirements for this function are addressed in the Technical Requirements Manual .
BYRON - UNITS 1 &2 B 3.7.5 - 3 Revision +/-GB
Attachment 6a Braidwood Station, Units 1 and 2 NRC Docket Nos. 50-456 and 50-457 Proposed Technical Specifications Bases Changes (Mark-Up) for Information Only
Accumulators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)
The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 2 and 3).
The accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) .
LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA . Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the large break LOCA blowdown ohase of ....a t:GGA , the ECCS acceptance criteria of and the recovery 10 CFR 50 .46 (Ref./ 4) could be violated.
phase of a small For an accumulator to be considered OPERABLE, the isolation break LOCA valve must be fully open with power removed, a contained volume~ 31% and~ 63% (6995 gallons to 7217 gallons) with a boron concentration~ 2200 ppm and~ 2400 ppm, and a nitrogen cover pressure~ 602 and~ 647 psig, must be met.
APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure
> 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.
This LCO is only applicable at pressures> 1000 psig. At pressures~ 1000 psig, the rate of RCS blowdown is such that the 10 CFR 50.46 the ECCS pumps can provide adequate injection to ensu re that peak clad temperature remains belm,* the 10 CFR 50 .46 (Ref. 4) acceptance CRef . 4) l i mit of 2200°F.
criteria are met BYRON - UNITS 1 &2 B3.5.1-6 Revision -18
Accumulators B 3.5 .1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed . The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.5 Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 1% volume increase (nominally 70 gallons or 10% of indicated level) will identify whether inleakage has caused a reduction in boron concentration to below the required limit . It is not necessary to verify boron concentration of the accumulator after a 1% volume increase (10% indicated level increase) if the added water inventory is from the Refueling Water Storage Tank (RWST) and the boron concentration of the RWST is~ 2200 ppm and~ 2400 ppm.
With the water contained in the RWST within the boron concentration requirements of the accumulators , any added inventory would not cause the accumulator's boron concentration to exceed the limits of this LCO .
With the only indication available to the operators in the control room being level indication in percent, a required accumulator volume increase of 1% or an increase of 10% of indicated level would require the accumulator to be sampled to verify the accumulator boron concentration is within the limits. The small break LOCA analysis assumes a nominal
.~ater volume of 7106 gallons based on 3the TS minimum and
~ mum__limi ~ -e.f. 6995 gallons 3(935 ft , 31% of indicated
~ 21. 7 gallons (965 ft , 63% of indicated level ).
These volumes are also indicated in the specific tank curves for the SI accumulators. The large break LOCA analysis assumes a water volume range of 6882 gallons (920 ft 3 , 15%
of indicated level) to 7331 gallons (980 ft 3 , 79% of small break and indicated level) which bounds the TS limits. The 10%
large break LOCA indicated level increase is considered a conservative analyses assume indication for a 70 gallon increase in the accumulator volume requiring an increase in the sampl ing requirement to verify accumulator boron concentration remains within the specified limits.
BYRON - UNITS 1 &2 B 3.5.1 - 9 Revi sion +5-
Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.6 Verification that power is removed from each accumulator isolation valve operator ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve . If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA.
The power to the accumulator motor operated isolation valves is removed by opening the motor control center breaker and tagging it out administratively. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. IEEE Standard 279-1971.
2. UFSAR, Chapter 15.
3. UFSAR, Chapter 6.
4. 10 CFR 50.46.
5. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
BYRON - UNITS 1 &2 B 3.5.1 - 10 Revision ::JI::r
ECCS-Operating B 3.5.2 BASES APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria for the ECCS, established by 10 CFR 50.46 (Ref. 2),
will be met following a LOCA:
During a small break LOCA maximum fuel element cladding temperature is~ 2200°F and during a large There is a high break LOCA there must be a high level of probability level of probability that maximum fuel element cladding temperature is s 2200°F; that the maximum Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation; C. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount that There is a high would be generated if all of the metal in the cladding level of probability cylinders surrounding the fuel, excluding the cladding that the maximum surroundi ng the plenum volume, were to react;
d. Core is maintained in a coolable geometry; and
e. Adequate long term core cooling capability is maintained.
The LCO also l imits the potential for a post trip return to power following an MSLB event and ensures that containment temperature limits are met.
Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Ref. 3). This event establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The centrifugal charging pumps and SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The SGTR and MSLB events also credit the centrifugal charging pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions: ~
a. For the large break LOCA event, the ASTRUM methodology examines LOOP and no-LOOP cases with a single failure disabling one train of SI pumps . No failure of containment heat removal system is modeled; and BYRON - UNITS 1 &2 B 3.5 .2 - 4 Revision -74
RWST B 3.5.4 BASES APPLICABLE SAFETY ANALYSES (continued)
In the ECCS analysis , the containment spray temperature is a range of RWST ~ ~ assumed to be equal to the RWST l rn,er temperature l iFAi t of temperatures of ~ If the lower temperature l imit is violated , the 32°F to 120°F is containment spray further reduces containment pressure . The used for the reduced containment pressure lowers the quality of steam containment spray exiting the break thus decreasing the rate which the steam is vented to the conta i nment atmosphere . The decreased rate temperature per of steam vented to the containment atmosphere results in a approved methods corresponding decrease in the rate the Reactor Coolant (Ref. 4). System pressure drops and the rate ECCS f l uid is injected in the core thereby causing a rise in peak clad temperature .
+fie-upper temperature limit of l00°F is used in the small break LOCA analysis and containment OPERABILITY analysis .
Exceeding this temperature will result in a higher peak clad temperature , because there is less heat transfer from the core to the injected n1ater for the small breal< LOCA and 1
higher containment pressures due to reduced containment spray cool ing capacity. For the containment response following an MSLB , the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.
The limits on RWST level and boron concentration also ensure that the post-LOCA sump pH will be between 8.0 and 11 .0.
The minimum and maximum pH values are verified for each fuel cycle using conservative maximum and minimum RWST volumes and the maximum and minimum allowed RWST boron concentrati ons . The LOCA offsite dose ana lysis assumes a conservatively low sump pH for the re-evolution of iodine from the sump . Ensuring that the minimum sump pH is at least 8.0 protects mechanical components and equipment inside containment from the effects of chloride induced stress corrosion cracking. Ensuring that the maximum sump pH is no greater than 11 .0 limits the production of hydrogen due to the corrosion of aluminum and zinc inside containment. Finally, the limits on RWST boron concentration also ensure that the containment spray pH is acceptable . The calcul ation of the iodine removal effectiveness of the containment spray assumes a conservatively low containment spray pH.
The RWST satisfies Criterion 3 of 10 CFR 50 .36(c)(2)(ii) .
The small break and large break LOCA analyses use a range of RWST temperatures of 32°F to 120°F per approved methods (Ref. 4) .
'--- -------lExceeding the maximum temperature of this range could result in a higher peak cladding temperature, because there is less heat transfer from the core to the injected water. The containment BYRON - UNITS 1 &2 B 3.5.4 - 5 Revis ion B
RWST B 3.5.4 BASES REFERENCES 1. WCAP-13964, Revision 2, "Comnonwea l th Edi son Company, Byron/Braidwood Units 1 &2, Increased Steam Generator Tube Plugging/Reduced Thermal Design Flow/Positive Moderator Temperature Coefficient Analysis Program, Engineering/Licensing Report," September 1994.
2. UFSAR, Chapter 15.
3. UFSAR, Section 6.2. 1.
4. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) ," November 2016 .
BYRON - UNITS 1 &2 B 3.5.4 - 9 Revision .g.
Attachment 6b Byron Station, Units 1 and 2 NRC Docket Nos. 50-454 and 50-455 Proposed Technical Specifications Bases Changes (Mark-Up) for Information Only
F0 (Z)
B 3.2.1 BASES BACKGROUND (continued)
Core monitoring and control under non-equilibrium conditions are accomplished by operat ing the core within the limits of the appropriate LCOs, incl uding the limits on AFD, QPTR (only when PDMS is inoperable), and control rod insertion.
APPLICABLE This LCD precludes core power distributions that violate SAFETY ANALYSES the following fuel desi gn criteria:
a. During a small break Loss Of Coolant Accident (LOCA) the peak cladding temperature must not exceed 2200°F 10 CFR 50.46 and during a large break LOCA there must be a high level of probability that the peak cladding acceptance criteria temperature does not exceed 2200°F (Ref. l);
must be met
b. During a loss of forced reactor coolant flow accident ,
there must be at least 95% probability at the 95%
confidence level (the 95/95 Departure from Nucleate Boiling (DNB) criteri on) that the hot fuel rod in the core does not expe rience a DNB condition;
c. During an ejected rod accident, the prompt energy deposition to the fue l must not exceed 200 cal/gm (Ref. 2); and
d. The control rods must be capable of shutting down the reactor with a min imum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).
Limits on F0(Z) ensure that the value of the initial total peaking factor assumed in the accident ana lyses remains valid. Other criteria must also be met (e .g., maximum cladding oxidation , maximum hydrogen generation , coolable geometry , and long term cooling) . Hmvever , the peak cladding temperature is typically most limiting .
FQ(Z) l imits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the F0 (Z) limit assumed in safety analyses for other postulated accidents.
Therefore, this LCO provides conservative limits for other postulated accidents.
F0(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
BYRON - UNITS 1 &2 B3.2.1-2 Revision -H F0 (Z)
B 3.2.1 BASES LCO The Heat Flux Hot Channel Factor, Fa<Z), shall be limited by the following relationships:
F RTP Fa (Z) :$; _a_ K(Z) for P > 0.5 p
F RTP Fa (Z) :$; _a_ K(Z) for P :$; 0.5
0.5 where
F~rP is the Fa(Z) limit at RTP provided in the COLR, K(Z) is the normalized Fa(Z) as a function of core height provided in the COLR, and p = THERMAL POWER RTP For this facility, the actual values of F'61P and K(Z) are given in the COLR; however, F'61P is normally a number on the
~ order ~ :-§9. , and K(Z) is a function that looks like the
~ ovided i n Figure B 3.2.1-1 .
Fa(Z) is approximated by Fg (Z) and Fi (Z). Thus, both Fg (Z) and Fi <z) must meet the preceding limits on Fa( Z).
When PDMS is inoperable, an Fg(Z) evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value <F;<z)) of F0( Z). Then ,
Fg(Z) = F;(Z) * (1. 0815) where 1.0815 is a factor that accounts for fuel manufacturing tol erances and flux map measurement uncertainty .
Fg (Z) is an excellent approximation for Fa( Z) when the reactor is at the steady state power at which the incore flux map was taken .
BYRON - UNITS 1 &2 B 3.2 .1 - 3 Revision -+/
F0(Z)
B 3.2 .1 BASES LCD (continued)
When PDMS is OPERABLE , Fq(Z) is determined continuously .
Then ,
F5 (Z) = Ft (Z)
UFO where Um is a factor that accounts for measurement uncerta inty (Ref . 4) and engineering uncertainty defined in the COLR.
The expression for F;(Z)is :
F~(Z) = F5(Z)
W(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation . W(Z) is included in the COLR. When PDMS is inoperable , the F5(Z) is cal culated at equilibrium conditions .
The F0 (Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during a small break LOCA and assures Hi th a high ensure that the 10 level of probability that the peak cladding temperature does not exceed 2200°F during a large break LOCA (Ref. 1).
CFR 50.46 acceptance criteria This LCD requires operation within the bounds assumed in the are met safety anal yses. Calculations are performed in the core design process t o confi rm that t he core can be controlled in such a manner during operation that it can stay within the LOCA F0 (Z) limits . If F5(Z) cannot be maintained within the LCO limits, reduction of the core power is required.
Violating the LCO limits for F0(Z) may produce unacceptable consequences if a design basis event occurs while F0 (Z) is outside its specified l imits.
APPLICABILITY The F0(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses . Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.
BY RON - UNITS 1 &2 B 3. 2.1-4 Revision 8-
F0 ( Z)
!This figure is replaced with the figure on the next page. I B 3.2 .1 1.2 r--** .. :* ...... *,**-***. **** **1. ;****--*-- --r-**--*-*-:
1 DO NOT OPERATE IN THIS AREA I
I I :
i (6.0, / 1.o) i ! !
1.0 ~1-------+-~~~-------r-1-...:.. . - . t - --*- ---- *-+ - - .- ..- -- ,
i ! ._j i
I I I
( 12 .0, 0 .925)
I I I I . .. . _j _ -- ---1 O. B I I
~~ Ir---.+--*
0. 6 I I
. -* .---* r --*l -.- .---* *- .
I I i .
I
0. 21- -
ooi~~~~~R REXlfu~e~~liR -r ...i 1**
1 - * -- r* I I
0t-- -*-*-*- -4* *- ---i--*---- -~---*-**-}o--*-/2 CORE HE IGHT (FT .)
Figure B 3.2.1-1 (page 1 of 1)
K(Z) - Normalized F0 (Z) as a function of Core Height BYRON - UNITS 1 &2 B 3.2.1 - 13 Revision 15
1.2 DO NOT OPERATE IN THIS AREA (0.0, 1.0) (6.0, 1.0) ( 12.0, 1.0) 1.0 0.8
~
~0.6 0.4 THIS FIGURE FOR ILLUSTRATION ONLY.
DO NOT USE FOR 0.2 OPERATION.
REFER TO COLR FOR ACTUAL VALUES.
0.0 0 2 4 6 8 10 12 CORE HEIGHT (FT.)
Figure B 3 . 2 . 1-1 (page 1 of 1)
K(Z) - Norma l ized F 0 (Z) a s a function of Core Height
BASES BACKGROUND (continued)
The COLR provides peaking factor limits that ensure that the design criterion for the Departure from Nucleate Boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. Al l DNB limited transient events are assumed to begin with an F~H value that satisfies the LCO requirements .
Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible claddi ng perforation with the release of fission products to the reactor coolant.
APPLICABLE Limits on F~H preclude core power distributions that exceed SAFETY ANALYSES the following fuel design limits:
a. There must be at least 95% probability at the 95%
confidence level (the 95/95 DNB criterion) that the hottest fuel rod i n the core does not experience a DNB condition;
b. During a small break Loss Of Coolant Accident (LOCA)
Peak Cladding Temperature (PCT) must not exceed 2200°F and during a large break LOCA there must be a high the 10 CFR 50.46 level of probability that PCT does not exceed 2200°F; acceptance criteria c. During an ejected rod accident, the prompt energy must be met deposition to the fuel must not exceed 200 cal/gm (Ref.3) ( Ref. 1) ; and ct. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a minimum required Shutdown Ma rgin with the highest worth control rod stuck fu ll y withdrawn.
For transients that may be DNB limited , F~H is a significant core parameter. The limits on F~H ensure that the DNB design criterion is met for normal operation, operational transients , and any transients arising from events of moderate frequency. Refer to the Bases for LCO 3.4 .1, RCS11 Pressure, Temperature, and Flow DNB Limits, for a 11 discussion of the applicable Departure from Nucleate Boiling Ratio (DNBR) limits.
BYRON - UNITS 1 &2 B 3.2.2 - 2 Revision 18
BASES APPLICABLE SAFETY ANALYSES (continued)
The allowable F1H limit increases with decreasing power level. This functionality in F1H is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1 .1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of F~H in the analyses . Likewise, all transients that may be DNB limited are assumed to begin with an initial F1H as a function of power level defined by the COLR limit equation .
The LOCA safety analysis indirectly models~!+ as an input parameter . The Nuclear Heat Flux Hot Channel Factor (FG(Z))
The Nuclear and the axial peaking factors are inserted directly into the Enthalpy Rise Hot LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature (Ref . 3) .
Channel Factor (FN~H), the Nuclear The fuel is protected in part by Technical Specifications, Heat Flux Hot which ensure that the initial conditions assumed in the Channel Factor safety and accident analyses remain valid. The following LCOs ensure this: LCD 3.1.6, "Control Bank Insertion (F 0 (Z)), and the Limits," LCD 3.2.1, "Heat Flux Hot Channel Factor (F0 (Z )),"
axial peaking LCD 3.2.2, LCD 3.2.3 , LCD 3.2 .4, and LCD 3.2 .5, "Departure factors are from Nucleate Boiling Ratio (DNBR) . "
supported by the
~H and ~(Z) are measured periodically using the movable LOCA safety analyses that verify incore detector system when PDMS is inoperable.
Measurements are generall y taken with the core at, or near, compliance with the steady state conditions . Core monitoring and control under 10 CFR 50.46 trans i ent cond i tions (Condition 1 events) are accomplished acceptance criteria by operating the core within the limits of the LCOs on AFD, (Ref.3) QPTR, and Control Ban k Insertion Limits . When PDMS is OPERABLE, FNH and F0(Z) are determined continuousl y. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on DNBR and Control Bank Insertion Limits.
F1H satisfies Criterion 2 of 10 CFR 50.36(c)(2)Cii).
BYRON - UNITS 1 &2 B 3.2 .2 - 3 Revision fl
BASES LCO F1H shall be maintained within the limits of the relationship provided in the COLR.
The F~H limit identifies t he coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB .
The limiting value of F~H* described by the equation contained in the COLR, is the design radial peaking factor used in the plant safety analyses.
The power multiplication factor in this equation provides margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of F1H is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER.
APPLICABILITY The F1H limits must be maintained in MODE 1 to prevent core power distributions from exceeding the fuel design limits the 10 CFR 50.46 for DNBR and -Pff.. Applicability in other modes is not required bet ause there is either insufficient stored energy acceptance criteria in the fuel or insufficient energy being transferred to the (Ref. 3) reactor coolant t o require a limit on the distribution of core power. Speci fically , the design bases events that are sensitive to F~H in other modes (MODES 2 through 5) have significant margin to DNB , and therefore, there is no need to restrict F~H in these modes.
ACTIONS A.1, A.2. A.3, and A.4 With F~H exceeding its limit, Condition A is entered. F~H may be restored to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , through, for example, realigning any misaligned rods or reducing power enough to bring F~H within its power dependent limit. If the value of F~H is not restored to within its specified limit, THERMAL POWER must be reduced to
< 50% RTP in accordance with Required Action A.1. When the F~H limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the F1H value (e.g., static control rod misalignment) are considered in the safety analyses.
BYRON - UNITS 1 &2 B 3.2.2 - 4 Revision +/
QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)
BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses . Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. When Power Distribution Monitoring System (PDMS) is OPERABLE , Peak Linear Heat Rate and the linear power along the fuel rod with the highest integrated power are measured continuously.
APPLICABLE Limits on QPTR preclude core power distributions that SAFETY ANALYSES violate the following fuel design criteria:
a. During a small break Loss Of Coolant Accident (LOCA )
the Peak Cladding Temperature (PCT) must not exceed 2200°F and during a large break LOCA there must be a high level of probability that the PCT does not exceed 10 CFR 50.46 2200°F (Ref . 1) ;
acceptance criteria must be met b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95%
confidence level (the 95/95 Departure from Nucleate Boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition ;
c. During an ejected rod accident, the prompt energy deposition to the fuel must not exceed 200 cal/gm (Ref . 2); and BYRON - UNITS 1 &2 B 3.2.4 - 1 Revision -l-8
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY ANALYSES, more ESFAS Functions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident. An ESFAS APPLICABILITY Function may be the primary actuation signal for more than one type of accident. An ESFAS Function may also be a secondary, or backup, actuation signal for one or more other accidents. For example, Pressurizer Pressure-Low is a primary actuation signal for srnall Loss Of Coolant Accidents (LOCAs) and a backup actuation signal for Steam Line Breaks (SLBs) outside containment. Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit.
These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions may also serve as backups to Functions that were credited in the accident analysis (Ref. 3).
The LCO requires all instrumentation performing an ESFAS Function to be OPERABLE when the unit status is within the Applicability. Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions.
The LCO generally requires OPERABILITY of three or four channels in each instrumentation Function and two channels in each logic and ma nual initiation Function. The two-out-of-three and the two-out-of-four configurations allow one channel to be tripped during maintenance or testing without causing an ESFAS initiation. Two logic or manual initiation channels are required to ensure no single random failure disables the ESFAS.
BYRON - UNITS 1 &2 B 3.3.2 - 6 Revision B-
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES , LCO, and APPLICABILITY (continued)
The required channels of ESFAS instrumentation provide unit protection in the event of any of the analyzed accidents .
ESFAS protection functions are as follows:
1. Safety Injection Safety Injection (SI) provides two primary functions:
compl iance with the Primary side water addition to ensure maintenance 10 CFR 50.46 or recovery of reactor vessel water level (coverage of the active fuel for heat removal, acceptance criteria clad integrity, and for liFRiting peak clad (Ref. 17) teFRperature to< 2200°F) ; and
2. Boration to ensure recovery and maintenance of SOM.
These functions are necessary to mitigate the effects of High Energy Line Breaks (HELBs) both inside and outside of containment. The SI signal is also used to initiate other Functions such as:
Phase A Isolation ;
Containment Purge Isolation;
Reactor Trip;
Turbine Trip;
Feedwater Isolation ;
Start of Auxiliary Feedwater (AF) pumps;
Control room ventilation isolation; and
Enabling automatic switchover of Emergency Core Cooling Systems (ECCS) pump suction to containment sump .
BYRON - UNITS 1 &2 B 3.3 .2 - 7 Revision B-
ESFAS Instrumentation B 3.3.2 BASES REFERENCES 1. UFSAR, Chapter 6.
2. UFSAR, Chapter 7.
3. UFSAR, Chapter 15.
4. I EEE-279-1971.
5. Technical Requirements Manual.
6. WCAP-12523, "RTS/ESFAS Setpoint Methodology Study,"
October 1990.
7. WCAP-10271-P-A, Suppl ement 2, Rev. 1, June 1990.
8. WCAP-13632 Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," August 1995.
9. UFSAR, Section 7.3.
10. WCAP-12583, "Westinghouse Setpoint Methodology For Protection Systems, Byron/Braidwood Stations," May 1990.
11. WCAP-14036-P -A , Revision 1, "Elimination of Periodic Protection Channel Response Time Tests ," October 1998.
12. WCAP-13877, Revision 2-P, "Re 1i abi l i ty Assessment of Westinghouse Type AR Relays Used as SSPS Slave Relays,"
October 1999.
13. WCAP-13878-P, Revis ion 2, "Reliability Assessment of Potter &Brumfield MOR Series Relays," October 1999.
14. WCAP-13900, Revis ion O, "Extension of Slave Relay Surveillance Test Intervals," April 1994.
15. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," October 1998.
16. WCAP-15376-P-A, Revision l, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times,"
March 2000.
117. 10CFR50.46 BYRON - UNITS 1 &2 B 3.3.2 - 58 Revision 62
Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5 .1 Accumulators large break BASES BACKGROUND The functions of the ECCS accumulators are to suppl ater to the reactor vessel during the blowdown phase of a Loss Of Coolant Accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a smal l break LOCA .
The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere. large break In the refill phase of a CA, which immedi ately follows t he blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available t help fill voids in the lower plenum and reactor vessel do ncomer so as to establish a recovery level at the bottom of the
-pn-it-ia_t_e~I core a~ ongoing reflood of the core with the additi n of Safety Injection (SI) water .
The accumulators are pressure vessel s partially fill d with borated water and pressurized with nitrogen gas. Th accumulators are passive components, since no opera r or control actions are required in orde r for them top rform their function. Internal accumulato r tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pre sure.
Initial accumulator inventory which is injected into the reactor vessel is lost out the break.
BYRON - UtiITS 1 &2 B 3.5.1 - 1 Revision B

v t e Letter Sequence Request
EPID:L-2020-LLA-0242, Proposed License Amendment Request, Addition of Analytical Methodology to the Core Opera Ting Limits Report for a Full Spectrum Loss of Coolant Accident (Fsloca) Gamma Energy Redistribution Information (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Administration Withholding Request Acceptance Questions 15 April 2021 Answers 20 May 2021 Results Approval Other: ML20063L282, ML21297A007
MONTHYEAR2021-10-19 [Table View]

RS-20-010, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

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RS-20-010, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

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3.0 TECHNICAL EVALUATION

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