BYRON 2018-0109, 10 CFR 50.59 Summary Report

10 CFR 50.59 Summary Report

ML18344A074
Person / Time
Site:	Byron
Issue date:	12/10/2018
From:	Kanavos M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BYRON 2018-0109
Download: ML18344A074 (57)

Text

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Byron Generating Station

_EEr ExelonGeneration 445ONorthGermanChurchRd www.exeloncorp.com December 10, 2018 10 CFR 50.59(d)(2)

LTR: BYRON 2018-0109 FILE: 1.10.0101 (YD.101,5A.109) 2.07.061 0 (5A. 109) 2.07.0611 (1D.101)

United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. SIN 50-454 and STN 50-455

Subject:

10 CFR 50.59 Summary Report Pursuant to the requirements of 1 0 CFR 5059, Changes, tests and experiments, paragraph (d)(2), Byron Station is providing the required report for Facility Operating License Nos. NPF-37 and NPF-66. This report is provided for the evaluations implemented for the time period of January 1 201 6 through November 30, 201 8 and consists of 10 CFR 50.59 Review Coversheets for changes to the facility or procedures as described in the Updated Final Analysis Report (UFSAR) and test or experiments not described in the UFSAR.

Please direct any questions regarding this submittal to Mr. Douglas Spitzer, Regulatory Assurance Manager, at (81 5) 406-2800.

Respectfully, Mark E. Kanavos Site Vice President Byron Generating Station MEKIGC/rm

Attachment:

Byron Station 10 CFR 50.59 Summary Report cc: NRC Regional Administrator NRC Region Ill

ATTACHMENT Byron Station 10 CFR 5O59 Summary Report

1. Evaluation No. Rev Title 1 6G-i 5-001 ECCSICS Hydraulic Analysis Incorporate Instrument Uncertainty from CS 0

Pump Testing Replace RCP Under frequency KF Relays with Circuit Shield Type 81 2 6G-1 6-004 0 Frequency Relays to Address ABB Part 21 Notification of Potential Defect For KF Relay ZPA (Unit 1) 3 6G-1 6-005 0 OBOA PRI-7 Loss of Ultimate Heat Sink Unit 0 (Revision 2)

Temporarily Defeat FW Water Hammer Prevention System (WHPS) FW 4 6G-1 6-007 0 Isolation Signals During Normal Power Operation for Steam Generators 2N2B/2C/2D APPENDIX J Scope Reduction as a Result of True North Engineering 5 6G-16-008 0 Report Commissioned by Corporate Engineering 6 6G-1 7-001 0 Lost Spent Fuel Pool (SFP) Crimps and RVLIS Pins Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet 7 6G-1 7-002 0 (N-i Outage) 8 6G-1 7-003 0 Westinghouse Ovation Digital Upgrade for DEHC 9 6G-17-004 0 TCCP to Install Pneumatic Jumper to Open 1WG074 Provide Capability to Disable Water Hammer Prevention System (WHPS)

Feedwater (FW) Isolation Signals During Normal Power Operation &

10 6G-17-005 0 Change WHPS SG Low Pressure/Level Control Relay Operation from Actuate On De-Energized To Actuate On Energized Process TORMIS LAR & Supporting Docs, TRM 3.7.e Tornado Design 11 6G-1 7-006 0 Basis SXCT Fans Operating, TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown.

6G-17-007 0 Lost Parts Evaluation for Head Lift Rigging j_

Evaluation No. 6G-15-OOJ

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of 7 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 392420 and DRP 17-034 Revision Number: QQ

Title:

ECCS/CS Hydraulic Analyses incorporate Instrument Uncertaintyirom CS Pump Testing NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

EC 392420 implements Revisions 4 and 5A of design anaiysis BYRO6-O29/BRW.O6OOi6-M. This analysis calculates flow rates for the Emergency Core Cooling System (ECCS) and Containment Spray (CS) System in support of the Braidwood and Byron Station design basis and accident analyses.

EC 392420 also includes design basis analyses that have been revised to incorporate the new ECCS and CS flow rates. The complete list of analyses that have been revised is as follows:

1. BYRO6-029/BRW-06-0016-M, S1..RHR-CS-CV System Hydraulic Analysis in Support of GSI-191, Revision 4 and Revision SA

2. ATD-O111, Maximum Containment Flood Level, Revision 15

3. SI9O4)1 Revision, Minimum Containment Flood Level, Revision 11

4. BRWO5-O63-M/BYRO5.O6i, G51491 Evaluation of Long Term Downstream Effects, Revision 4

5. BRW-06..0035-MJBYRO6-058, NPSIIA for RUR and CS Pumps During Post..LOCA Recirculation, Revision 3

6. BRW-97-0337-M1BYR97491, Containment Spray System Hydraulic Model, Revision 3

7. BRW-06-0028-MJBYRO6-030, Post-LOCA Chemical Effects Analysis in Support of GS1491, Revision 3

8. BYR97-4O6/BRW-97-O965M, RCFC Performance Curve Calculation, Revision 1

9. S1TH-1, Refueling Water Storage Tank (RWST) Level Setpoints, Revision 8

10. CAE..07..49/CCE-07-48, Phase 2 Evaluation of Reduced SI flow During Recirculation Phase of ECCS, Revision 1A.

11. CN.CRA-O7-47, ByronlBraidwood Unit 1 Steam Generator Tube Rupture Margin to Overfill, Revision 4

12. CN-CRA-07-72, Byron/Braidwood Unit 2 Steam Generator Tube Rupture Margin to Overfill, Revision 3

13. CNCRA49..67, ByronlBraidwood Unit 2 Steam Generator Tube Rupture Input to Dose, Revision 1

14. CN-CRA-1O-29, Byron and Braidwood Units I and 2 Main Steamline Break It/OC Evaluation of M&E Releases and Containment/Compartment Responses, Revision 1

15. CN.CRA-1O-30, Byro&Braidwood Unit I Steam Generator Tube Rupture Input to l)ose Analysis, Revision I

16. CN..CRA4O-54, LOCA Mass and Energy Release and Containment Integrity Analysis, Revision 2

17. CN-CRA-13-5, Byron and Braidwood Units 1: Steamline Break Mass and Energy Releases Outside Containment and Compartment Response, Revision 1

18. CNCRA-13-6, Byron and Braidwood Units 2: Steamline Break Mass and Energy Releases Outside Containment and Compartment Response, Revision I

19. CN-CRA43-21, Byron and Braidwood Unit 2 Steamline Break Mass and Energy Releases inside Containment, Revision 0

20. CN-CRA-13-26, Byron and Braidwood Unit 2 Steamline Break Containment Response Analysis, Revision I

21. CNCRA43-29, Byron and Braidwood Unit 1 Stearnilne Break 11ass and Energy Releases Inside Containment, Revision 0

Evaluation No. 6G-15-OO1 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of 7 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 392420 and DRP 17-034 Revision Number: fQ

Title:

ECCS/CS Hydraulic Analyses Incorporate Instrument Uncertainty from CS Pump Testing

22. CN-CRA4440, Byron and Braidwood Unit 1 Steamllne Break Containment Response Analysis, Revision 1

23. PSA-B-9848, ByronfBraidwood ECCS flow Calculations for Safety Analysis, Revision 4

24. LTR-PL-13-32, Byron/Braidwood Unit I and 2 Increased Delay Time in Auxiliary Feedwater Injection Final Engineering Report, Revision OO1A

25. 3 $A-096.018, Head Loss Calculation, Revision 9 Note: The common Byron/Braidwood accident analyses were revised in conjunction with a Braidwood change to the maximum UHS supply water temperature (Braidwood EC 396478). These analyses are conservative for Byron as they use inputs that are more limiting than the Byron maximum UHS temperature of 100 °F, described in UFSAR Section 9.2.1.2.1.

In addition, the impact on the Peak Clad Temperature (PCT) due to the revised ECCS flow rates has been evaluated to be an increase of 2 °F for the EBLOCA and 0 °F for the SBLOCA. The evaluations are documented in the following Westinghouse documents:

1. Letter LTR-LIS-14446, dated May 2, 2014

2. Letter LTR-LIS-14-237, dated October 22, 2014

3. Letter LTR.LI$.14-245, dated October 27, 2014 The PCT evaluations and results were incorporated into the IOCFR5O.46 Annual Report (Reference EC 401531).

DRP No.17-034 incorporates the revised design analyses results into the UFSAR for Byron. UFSAR Chapter 6.2 is revised to document the inputs and results of the revised analyses of containment response to postulated line breaks inside containment. The UF$AR Chapter 6.3 description of the minimum CS pump flow/CS spray nozzle flow is revised and the RH, CV, and SI pump performance curves are revised.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

Revision 4 to design analysis BYRO6-O29IBRW.06-0016-M incorporates instrument uncertainty applicable to the in-service testing for the CS pumps into the analysis. This is an item that was identified in Braidwood IRs 1048015, 1050763, and Byron IR 1049823. Braidwood received a NCV for this issue in their 5/6/2010, NRC Component Design Basis Inspection (CDBI) Report.

In addition to incorporating instrument uncertainty for the CS pumps into the flow analysis, Revision 4 of BYR06-029fBRW-O6.0016M also revised instrument uncertainty values for the CV, SI, and Ru pumps to be consistent with the values from calculation P$A-B-98-08. Revision 4 also analyzed the impact of lower RWST internal pressures on the CS pump degradation limits and on the minimum ECCS injection flow rates.

The new ECCS and CS flow results from BYR06-029fBRW06-0016-M, Revision 4, were then used as input to revise various accident analyses. The common Byron/Braidwood accident analyses were revised in conjunction with a Braidwood change to the maximum UHS supply water temperature (Braidwood EC 396478). The higher new Braidwood input on UHS supply water temperature was conservatively applied to the analyses for Byron.

Evaluation No. 6G-J 5-001 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 3 of 7 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 3942O and DRP 17-034 Revision Number: Q

Title:

ECCS/CS Hydraulic Analyses Incorporate instrument Uncertainty from CS Pump Testing For the LOCA mass and energy (M&E) analyses, changes were also made to address Westinghouse Nuclear Safety Advisory Letters (NASLs) NA$L-06-6, NASL-11-5, and NASL44-2. Additional corrections were made to the SATAN78 power shape selection option to select a chopped cosine power shape and to incorporate the WCAP-10325-P-A evaluation methodology for drift flux and break flow with inertia.

The UFSAR change incorporates the results from the revised analyses.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

This activity does not have any impact on the operation of ECCS and CS pumps during normal or accident plant operations. The revised ECCS and CS flow rates have been evaluated for impact on design basis and accident analyses; the results are within the specific acceptance criteria.

The hydraulic analysis determines an acceptable range of performance for the ECCS and CS pumps. To provide a range of performance, for the cases that use maximum pump performance, the performance curves of the ECCS pumps have been enhanced by 3% (10% for the RH pumps) and, for the cases that use minimum pump performance, the pump curves have been degraded by 10%. The performance curves for the CS pumps have also been enhanced by 3%, but the allowed degradation is limited to the value that has been found acceptable in the applicable safety analyses. The maximum allowed degradation for the CS pumps, Train A and B, has been determined in calculation BYRO6-O29IBRW-06-0016-M, Revision 5A. The acceptable performance ranges will be incorporated as acceptance criteria for the pump performance surveillance testing.

The impact on the PCT due to the revised ECCS flow rates has been evaluated to be an increase of 2 °f for the LBLOCA and 0 °F for the SBLOCA. The PCT evaluations and results were incorporated into the IOCFR5O.46 Annual Report.

The revised ECCS flows do not impact the calculated results for a postulated Steam Generator Tube Rupture event. The revised LOCA and Steamline Break inside containment analyses show that the peak calculated containment pressure is below the peak calculated containment internal pressure (Pa), as referenced in Byron Technical Specification 5.5.16. The revised containment temperature profile is bounded by the EQ temperature profile. For postulated steam line breaks outside of containment, the changes had little effect on the limiting transients and the current UFSAR results and figures remain valid.

The changes/corrections made to the LOCA M&E analysis method corrected Identified errors and incorporated NRC approved model options. The corrections associated with NASL-06-6, NASL4I-5, and NASL44-2 result in a penalty to calculated containment pressure and temperature.

The calculated minimum containment flood level remains greater than the minimum required flood level for operation of the Containment Recirculation sumps. The calculated maximum containment flood level remains less than the evaluated containment flood level described in the UFSAR.

The Net Positive Suction Head (NPSH) Available during recirculation operation remains greater than the NPSH Required for the ffl-I and CS pumps. in addition, the void fraction at the pumps remains within the pump acceptance criteria.

Evaluation No. 6G-15-OOJ 50.59 REVIEW COVERSHEET FORM LS-AA-104-lOOi Revision 4 Page 4 of 7 Station/Unit(s): Byron/Units 1 and 2 ActivitylDocument Number: C 392420 and DRP 17-034 Revision Number: Q

Title:

ECCS/CS Hydraulic Analyses Incorporate Instrument Uncertainty from CS Pump Testing The revised ECCS and CS flow rates were determined to have no impact on the RW$T level setpoints or the operator action times for manual switchover from injection to recirculation postLOCA.

ECCS sump downstream wear was re-evaluated. The components evaluated for wear show an acceptable level of wear for a mission time of 30 days. Pump performance remains adequate to meet the required post-LOCA flows.

The revised postLOCA chemical effects analysis shows that the vendor (CCI) chemical effects testing for the containment sump screens used bounding chemical quantities. The calculation changes do not change the performance of the containment sump screens.

The results of the ECCS/CS hydraulic analysis show that the minimum CS flow to Containment remains above the input for Iodine removal in the LBLOCA dose analysis. With the maximum allowed CS pump degradation (minimum CS pump flows) the CS spray nozzle flow will be lower than the UF$AR described minImum value of 15 gpm per nozzle, The lower spray nozzle flow was evaluated and the resultant drop size remains less than the 1,000 i limit used in the containment analysis. The small reduction in the spray nozzle flow could result in a reduction in the spray coverage area for each nozzle. The small reduction in spray coverage area for each nozzle will have an insignificant impact on the overall sprayed containment volume used in the dose analysis.

Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 5O59 Screening, 5ft59 Evaluation, or a License Amendment Request, as applicable, is not required.)

Based on the results of the 50.59 Evaluation, the activity proposed in EC 392420 can be implemented without a License Amendment Request.

EC 392420 and the associated UFSAR changes do not introduce the possibility of a change in the frequency of an accident because the calculations that are addressed in EC 392420 evaluate system performance for postulated accidents and are not associated with an initiator of any accident. No new failure modes are introduced due to the results of the calculations. The results of the calculations that use the revised flow rates are acceptable and support post-accident operation of the affected equipment. The UFSAR changes in DRP 17-034 will assure that analyses inputs and assumptions are, and continue to be, complied with.

The activity that is addressed in EC 392420 does not make changes to plant equipment as to change the likelihood of a malfunction of an SSC because the revised calculations evaluate postulated accidents and failures. No new failure modes are introduced. Adequate Net Positive Suction Head (NPSH) Available has been confirmed for the ECCS and CS pumps while taking suction from the Containment recirculatfon sumps (BRW-06-0035-MJBYRO6-058); therefore, this activity does not increase the likelihood of a malfunction of an SSC important to safety. Adequate NPSH from the RWST had already been verified in calculation BYRO4-016, Revision 3; this analysis was not affected by the revised ECCS/CS flow rates because the analysis uses runout flow rates. The changes that were made to the GSII91 Downstream Effects Analysis (BRW-05-063-M/BYRO5-061, Revision 4) do not introduce additional debris in the post-LOCA recirculated water or change the sump screens; therefore, the likelihood of a malfunction for ECCS and CS pumps is not increased.

Evaluation No. 6G-15-OO1 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 5 of 7 Station/Unit(s): Byron/Units I and 2 Activity/Document Number: EC 392420 and DRP 17-034 Revision Number: QQ

Title:

ECCS/CS Hydraulic Analyses Incorporate Instrument Unceainty from CS Pump icting The calculated minimum containment flood height remains greater than the minimum required flood height for operation of the containment recirculation sumps. The calculated maximum containment flood level remains less than the evaluated containment flood level described in UFSAR Section D3.6.L2. The revised containment temperature profile is bounded by the Environmental Qualification temperature profile used for qualification of EQ equipment. Thus, the proposed changes do not increase the likelihood of occurrence of a malfunction of equipment in;portant to safety inside containment.

The proposed changes had essentially no impact on the calculated pressure and temperature for postulated steam line breaks outside of containment. Thus, there is a less than minimal increase in the likelihood of malfunction of equipment impacted by a MSLB outside of containment.

The proposed changes do not change the RWST level setpoints, the required time to perform the manual actions to complete switchover from the post-LOCA injection mode to the recirculation mode of operation, or the switchover procedure steps. Thus, the likelihood of a malfunction during switchover is unchanged.

The results of the revised hydraulic analysis show that the Containment Spray flow rates bound the flow that was used in the LOCA dose analysis. The change has an insignificant impact on the input parameters (fraction of containment sprayed and containment volume sprayed) used in the LOCA dose analysis. The revised flow rates for the Steam Generator Tube Rupture event do not have an impact on the results of the dose analysis.

The calculated peak containment pressures are below the current Technical Specification 5.5.16 peak calculated containment pressure value, Pa, and containment design pressure. Thus, containment integrity is unchanged. Fuel integrity is unchanged because the calculated peak clad temperature remains below the design limit.

This activity does not affect the operation of plant equipment. The ECCS and CS pumps will be able to operate as credited in the UFSAR in response to a malfunction of an SSC. The revised CS flow rates have been evaluated and have been found to be higher than the flows used in the dose analyses and the CS spray nozzle drop size remains acceptable In addItion, the peak calculated containment pressure is below the Pa value in Technical Specifications 5.5.16. Therefore, this activity does not increase the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

EC 392420 does not create the possibility for an accident of a different type than any previously analyzed in the UFSAR. The calculations that are addressed in EC 392420 evaluate system performance for postulated accidents and are not associated with an initiator of any accident. No new failure modes are introduced.

The analyses that are implemented with EC 392420 do not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce a new failure result that is not bounded by those described in UFSAR Section 6.2.1.3.5 (LOCA), UFSAR Section 6.2.1.4.3 (MSLB), and UFSAR Table 6.5-1 (CS System). The revised analyses use limiting inputs based on the UFSAR described, design basis, postulated failures. These analyses show that the results are acceptable and bounded by existing acceptance criteria. Therefore, a malfunction of an SSC important to safety does not have a different result than any previously evaluated in the UFSAR.

The proposed activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered. The peak calculated containment pressure is less than the design basis Pa

Evaluation No. 6G-15-OO1 50.59 REVIEW COVERSHEET FORM LS-i1O4-1OO1 Revision 4 Page 6 of 7 StationlUnit(s): Byron/Units 1 and 2 Activity/Document Number: EC 392420 and DRP 17-034 Revision Number: Q

Title:

ECCS/CS Hydraulic Analyses Incorporate Instrument Uncertainty from CS Timp Testing as given in Byron rfechnical Specifications 5.5.16. The impact on the Peak Clad Temperature (PCT) due to the revised ECCS flow rates has been calculated to be an increase of 2 °F for the LBLOCA and 0 °F for the SBLOCA. The increase in PCT was included in EC401531, 2015 ANNUAL 10 CFR 50.46 REPORT:

BYRON AND BRAIDWOOD UNIT I & 2. PCT remains below the UFSAR described design basis limit.

The design analyses changes for EC 392420 do not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses. UF$AR Section D3.6d describes the analysis for maximum containment flooding level. The analysis method and assumptions are unchanged. The analysis does not credit water trapped in the refueling cavity consistent with the requirement of Regulatory Guide 1.82. The change to the minimum flood level analysis does not revise the amount of trapped water.

U?SAR Appendix A associated with Regulatory Guide 1.82, discusses that the post.LOCA chemical effects has been evaluated for the sump screens, however the method of evaluation is not specified in the UFSAR or in Regulatory Guide 1.82. The method of evaluation for downstream effects wear is not specified in the UFSAR. The method of evaluation for determining RH and CS pump available NPSH is unchanged. No credit is taken in the NPSH analysis for increase in containment pressure due to the accident. Consistent with Regulatory Guide 1.82 and the UFSAR description, no credit is taken in the NPSH analysis for increase in containment pressure due to the accident. The UFSAR described Computational Fluid Dynamics (CFD) analysis method for calculating containment sump head loss is unchanged. The method of evaluation for calculating RCFC performance is not specified in the UFSAR. UFSAR Section 6.3.5.4 describes the basis for the RWST level setpoints, however the method of evaluation is not specified in the UFSAR.

The revisions for the Steam Generator Tube Rupture (SGTR) Margin to Overfill and the SGTR input to dose analyses revise Inputs used in the analyses, and the UFSAR described method of evaluation is unchanged. The outside containment steam line break analyses were revised to evaluate the change in input for ECCS flow and higher UHS supply water temperature when used as the supply for the AF system. The method of evaluation used to determine the main steam tunnel compartn;ent response temperatures for equipment qualification (EQ) was not changed. The method of evaluation for calculating the main steam tunnel EQ temperatures is not specified in the UFSAR.

UFSAR Section 6.2.1.1.3 describes the method for calculating peak containment pressure and temperature.

The containment analyses use the COCO software to determine the containment temperature and pressure post-accident. The required CS droplet size in the COCO software is 1000 or less. The minimum CS flow rate for each Spray nozzle is 13.7 gpm for Train B. The resulting drop size is less than 1,000 p Thus, the change in CS minimum nozzle flow does not result in a departure from the method of evaluation as described in the UFSAR.

The corrections associated with NASL.06-6, NASL4I.5, and NSAL-14.2 were explicitly addressed in the revised analysis. A description of the issues related to the subject NASLs and the potential impact on the LOCA M&E analysis was documented in a RAI response letter for Bra.idwood (Letter RS45429 dated 413012015). The correction of errors to address the issues related to NASLs-06-6, 5, and 2 were reviewed and accepted as described in the NRC Safety Evaluation for Braidwood License Amendment No.

189. The analysis method and application accepted for Braidwood, is the same analysis method and application being used at Byron. Since the method used has been approved by the NRC for the intended application, the corrections associated with NASL-06-6, NASL-11.5, and NSAL-14-2 are not a departure from a method as described in the UFSAR.

Evaluation No. 6G-15-OOJ 50.59 REVIEW COVERSHEET FORM LSAA-1O4-1OO1 Revision 4 Page 7 of 7

$tationfUnit(s): Byron/Units 1 and 2 ActivitylDocurnent Number: EC 392420 and DRP 17-034 Revision Number: QL

Title:

ECCS/CS Hydraulic Analyses Incorporate Instrument Uncertainty from C&?ump Testing The $ATAN7$ input model was revised to include and credit the effects for drift flux and break with inertia consistent with the 2/17/87, NRC SER for WCAP40325P, Westinghouse LOCA M&E Release Model for Containment Design. In addition, the WCAP4O325P evaluation methodology has been approved to incorporate improved fluid niornentum flux terms, which were applied to the break flow model.

The $ATAN78 input model was revised to model a chopped cosine power shape so the exit temperature from the upper core node equals the core outlet temperature. This change is per the NRC approved LOCA M&E release methodology in WCAP4O32S-P.

rfhe calculation revisions for the inside containment steam line break analyses revised inputs to the analyses and the UFSAR described method of evaluation is unchanged.

All the 50.59 Evaluation questions were answered No. Thus, the activity may be implemented per plant procedures without obtaining a License Amendment.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

ci Applicability Review D 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 50.59 Evaluation No.

6G1 5-001 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated wIth the Activity.

Evaluation No. 6G-1 6-004 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of3 Station/Unit(s): Byron I Unit 1 Activity/Document Number: EC 402344/TS Bases Change 16-008 Revision Number: QQ

Title:

Replace RCP Underfreguency KF Relays with Circuit Shield Type $1 Frequency Relays to Address ABB Part 21 Notification ofPotential Defect for KF Relay ZPA NOTE: for 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description ofwhat the proposed activity involves.)

The proposed activity replaces the existing Reactor Coolant Pump (RCP) bus underfrequency relays on the four 6.9 kV buses.

There are currently 8 existing electromechanical relays (ABB Type 1(F), two on each bus for an A and B train logic actuation that will be replaced with 8 new solid-state relays (ABB Type 8 1). These relays provide input to the Reactor Protection System (RPS), where an underfrequency condition on two ofthe four 6.9 kV buses, determined by either logic train above P-7, produces a reactor trip and a trip ofthe four reactor coolant pumps.

The current RCP underftequency relays are powered from the RCP 6.9kV bus Potential Transformers (PT). The new solid state relays will require a DC power input. The power supply required will be installed in this modification.

The Technical Requirements Manual (TRM) currently describes the RCP underfrequency relay nominal setpoint as 57Hz. The new ABB Type 8 1 underfrequency relays being installed by this modification will have the same nominal setpoint. The Technical Specification 3.3.1 Allowable Value for the RCP underfrequency relay is 56.08 Hz; no change to the Technical Specification Allowable Value is required for the new ABB Type 8 1 underfrequency relays.

The proposed activity will also revise calculations to document numerous loading changes that have been added to the DC system. The calculations determine the voltage drops, required voltages at the safety-related DC components, battery performance for the designed duty cycle and battery charger performance.

. BYR97-204 125 VDC Battery Sizing Calculation This calculation is to verify that the ESF 125V DC batteries are adequately sized to supply the design duty cycle in both the normal and cross-tied alignments and ensure a minimum end ofduty cycle voltage

. BYR97-205 125 VDC Battery Charger Sizing Calculation This calculation is to demonstrate that the existing 400A battery chargers have sufficient capacity to perform their intended functions

. BYR97-224 125 VDC Voltage Drop Calculation This calculation is to demonstrate that all safety-related loads from the 125 VDC distribution centers can perform their intended safety functions at the system minimum voltage. This does not include a verification to the AC bus Due ripping and closing coils

. BYR 2000-136 Voltage Drop Calculation for 4160V Switchgear Breaker Control Circuits This calculation evaluates the adequacy ofthe voltage at the 4160V break trip and close coils.

. BYR 2000-191 Voltage Drop Calculation for 480V Switchgear Breaker Control Circuits This calculation evaluates the adequacy ofthe voltage at the 480V break trip and close coils.

. BYR 2000-136 Voltage Drop Calculation for 6.9kV Switchgear Breaker Control Circuits This calculation evaluates the adequacy ofthe voltage at the 6.9kVV break trip and close coils.

. BYR1 1-003 Calculation for 56 Cell operation ofBatteries 1DCO1E, 1DCO2E, 2DCO1E and 2DCO2E This calculation is to evaluate the capacity margin for operation with less than 58 cells for the unit 1 and unit 2 ESF 125VDC batteries at the end ofthe duty cycle.

Due to changes in the battery loading the TS bases B 3.8.4 DC Source-Operating will be revised to show the new minimum required voltage for the ESF batteries.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

The manufacturer ofthe existing underfrequency relays has issued a 10 CFR 21 notification (Event # 50691, dated 12/18/14) alerting affected licensees that a review ofthe seismic qualification ofthe devices determined that certain styles ofthe relays do not meet the previously published seismic rating. An engineering review, Operability Evaluation 15-001 (EC 400644), has concluded that the Byron underfrequency relays do not conform to station seismic requirements and should be replaced.

Evaluation No. 6G-16-004 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of 3 Station/Unit(s): Byron I Unit 1 Activity/Document Number: EC 402344/TS Bases Change 16-008 Revision Number: QQQ

Title:

Replace RCP Underfreguency KF Relays with Circuit Shield Type 81 Frequency Relays to Address ABB Part 21 Notification ofPotential Defect for KF Relay ZPA The battery calculations are being revised to incorporate additional loads to the 125VDC system, and to increase the margin between the calculated minimum 125V DC battery terminal voltage and the required minimum voltage.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

Like the existing relays, the new underfrequency relays will receive the bus frequency input from the existing 6.9 kV bus potential transformers and, on a detected low frequency input into the RPS, will initiate a reactor and RCP trip. The trip setpoint and time delay will remain the same. The new relays will be powered from 125 VDC ESF distribution panels, whereas the existing relays are powered only from the 6.9 kV bus potential transformers. The change in power supply to the relays introduces a new failure mechanism and is considered as adverse in the Screening.

The revised calculations listed above demonstrate that after the addition ofthe new DC loads the battery terminal voltage will be acceptable to support the operation ofthe safety-related loads. In addition the revised BYR1 1-003 Calculation for 56 Cell operation ofBatteries 1DCO1E, 1DCO2E, 2DCO1E and 2DCO2E demonstrates that with less than 58 cells the batteries will be able to meet the design duty cycle requirements .There is no impact on the design bases, UFSAR described methodology or safety analyses described in the UFSAR by the listed calculation revisions.

Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The proposed change replaces the existing RCP electro-mechanical underfrequency relays with new solid state underfrequency relays. The change is required because the existing relays do not meet the seismic requirements. This change will require a new 125V DC power supply to be provided for the new relays. The existing nominal setpoint and T.S allowable value will be retained.

The RCP underfrequency relays are required to provide input for a reactor trip and a trip of all four RCPs on a detected low frequency. The new relays do not change the method of performing these functions. The result of a detected low frequency remains unchanged.

The existing relays use the bus PTs for both frequency input and power. The new relays will continue to use the bus PTs for the frequency input; however, they will require a new 125 V DC power supply. This introduces a new failure mechanism: loss of DC power to an underfrequency relay. Ifthe DC power is lost the relay will fail to the not tripped condition. This provides a reduction in reliability of the relay to perform its function, and this is considered to be an adverse impact. The new failure mechanism was reviewed in the 50.59 evaluation and determined to result in no more than a minimal increase in the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR. This was due to the reliability of the DC power system, the fact that failure ofthe DC power supply is alarmed in the control room and the ability ofthe operators to establish the redundant DC power supply.

The methodology used in the calculation for determining the 125VDC Battery sizing to meet the design duty cycle follows the methodology described in the UFSAR for this determination. The methodology used in the calculation to determine the battery charger size follows the methodology described in the UFSAR.

Therefore, based on this Screening and Evaluation the activity may be implemented under the governing procedures.

Evaluation No. 6G-16-004 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 3 of3 Station/Unit(s): Byron I Unit 1 Activity/Document Number: EC 402344/IS Bases Change 16-008 Revision Number: QQQ

Title:

Replace RCP Underfreguency KF Relays with Circuit Shield Type 81 Frequency Relays to Address ABB Part 21 Notification ofPotential Defect for KF Relay ZPA Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

El Applicability Review 50.59 Screening 50.59 Screening No. 6E-16-073 Rev. 0 50.59 Evaluation 50.59 Evaluation No. 6G-16-004 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-16-005 50.59 REVIEW COVERSHEET FORM LS-AA-t04-1001 Revision 4 Page 1 of3 Station/Unit(s): Byron Unit 1/Unit 2 Activity/Document Number: OBOA PRI-7 Revision Number: 2

Title:

OBOA PRI-7 Loss of Ultimate Heat Sink Unit 0 NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of I 0 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

The activity is to revise OBOA PRI-7, Loss of Ultimate Heat Sink Unit 0, to allow the use of the non-safety related piping connection (1/2 fX307B-6) between the unit 1/2 B Essential Service Water (SX) pumps and the B Auxiliary feed Water (AFW) Shaft Driven Cooling Water Pumps (1/2 SXO4P) in the beyond-design-basis event where there is a complete loss of ultimate heat sink cooling capability. This flow path was installed under EC 393374 to support the requirements of the Flexible and Diverse Coping Strategies (FLEX) Implementation Guide, NEI 12-06 and NRC Order EA-l2-049, Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

Currently Guideline 1/2 BFSG-2 Alternate AFWIEFW Suction Source for Unit I and Unit 2 is approved to allow the use o the FLEX piping FX307B-6 to connect the discharge of the B SX pumps to the suction of the B AFW pump following a toss of all AC power if the CST is not available and an alternate suction source for the B AFW pump is required. In order to enable the probabilistic risk assessment (PRA) evaluations for Byron Station to credit this connection during a beyond-design-basis event ofa loss ofall SX, a station procedure is required to implement the actions to connect the SX alternate flow path to the AF Shaft Driven Cooling Water Pump. The Operating Abnormal Procedure OBOA PRI-7 currently exists to address a complete loss of ultimate heat sink cooling capability; however, it does not currently credit the use of the FLEX alternate flow path connection piping FX307B.

Currently, upon a loss of all SX pumps the B AFW Shaft Driven Cooling Water pump takes suction from the SX system, pumps it through the B AFW pump room cooler, engine cooler, and gear coolers to maintain operation of the B AFW pump.

However the discharge of the Shaft Driven Cooling Water pump cycles back into the suction of the pump due to the hydraulic pressures in the SX system. This creates a cycling effect where the cooling water being used by the B AFW pump is being heated by the waste heat from the pump. The current design documents demonstrate the B AFW pump can operate for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in this condition. For extended operation of the B AFW pump following this beyond-design basis event (a complete loss of ultimate heat sink cooling capability) a new cooling source is needed. This activity, revising OBOA PRI-7, provides guidance on aligning the FLEX alternate path to provide a new cooling source. When the FLEX alternate path is aligned, the Shaft Driven Cooling pump will take suction for the Essential Service Cooling Water Basin and discharge back to the basin, thus ensuring long term access to sufficient cooling water to support the extended operation of the B AFW pump.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

The eftct of this activity is to allow alignment the non-safety related FLEX alternate flow path between the SX pump discharge and the B AFW Shaft Driven Cooling water pump suction following a complete loss of ultimate heat sink cooling capability.

This allows for extended, greater than 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, run times for the B AFW pump in the condition where there is no forced SX flow on either Unit. The AFW system has a safety related design function to operate during a loss of all AC a condition which also results in a loss offorced SX flow. The time period for B AFW pump operation during a loss ofall AC is described in the UFSAR as being a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Connecting the non-safety related FLEX piping to the safety-related AFW system following a complete loss of ultimate heat sink cooling during this 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period adversely affects the AFW system. The non-safety related FLEX piping has been designed (as described in EC 393374) as Augmented Quality, seismic category I and is designed to remain functional after a Design Basis Earthquake. There the piping is considered no likely to fail during the maximum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> operation it will be required to support the AFW pumps design requirement. In addition a failure of the non-safety related piping such that it can no longer support operation of the B AFW pump is prevented from damaging the pump due to the existing high temperature trips on the pump.

Evaluation No. 6G-1 6-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-100I Revision 4 Page 2 of 3 Station/Unit(s): Byron Unit 1/Unit 2 ActivitylDocument Number: OBOA PRI-7 Revision Number: 2

Title:

OBOA PRI-7 Loss of Ultimate Heat Sink Unit 0 Therefore, the safety related piping connection remains available if needed to replace a degraded non-safety related FLEX piping.

The Operator will make the determination that the condition where there is no SX forced flow will exist for longer than an hour, and if it will then the FLEX piping will be used, thus ensuring the safety related piping remains available to support pump operations for as long as practical.

Summary of Conclusion for the Activitys 50.59 Review:

(Providejustification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The design requirement of the AFW system, for the loss ofall AC event, is to provide a safety related injection to the steam generators for a period of at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This activity permits the use of a non-safety related piping connection to the AFW system following a beyond-design-basis event (complete loss of ultimate heat sink cooling capability) to allow extending the operation time of the AFW system beyond the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> requirement, however this piping system may be connected prior to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> operation time being exceeded.

The piping has been designed to withstand a design basis seismic event and continue to function. The water source for the Shaft Driven Cooling pump will be changed from the outlet of the SX strainer, to the inlet of the strainer, and the effects of this non-strained water has been evaluated to not degrade the AFW pump performance. The use of procedure OBOA PRI-7 to direct the evolution to connect the FLEX alternate piping has been reviewed and determined to be within the allowed use for an Operating Abnormal Procedure as described in UFSAR section 1 3.5 Plant Procedures. However, the Af system currently is designed to start, initiate and provide injection flow to the Steam Generators without operator intervention. The change to OBOA PRI-7 requires operator intervention, albeit an optional intervention, that has been determined to be adverse in the screening.

In addition allowing the connection of the non-safety related FLEX piping to the AFW system during the time period where the AFW system is described as performing a safety related function is determined to be adverse in the screening, and will be reviewed in the 50.59 Evaluation.

The attached Evaluation has demonstrated that the effect of using the non-safety FLEX piping does not present a more then minimal increase in the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The use of the non-safety related seismic Flex piping, as been determined to be adverse in the attached screening. This adverse condition was determined to be a not more than minimal increase in the likelihood ofa malfunction ofan SSC. In addition the operator intervention provided by the procedure has been demonstrated to not be more than a minimal increase in the likelihood of a malfunction ofan SSC due to the option to use operator action to extend the operational time for the AFW system during a toss of all forced SX flow.

Therefore, based on this Screening and Evaluation the activity may be implemented under the governing procedures.

This activity is determined to be adverse, based on connecting a non-safety related piping section to the AFW during the time period (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> runtime requirement) where the AFW system is performing a safety related activity during a loss ofall AC.

Attachments:

Attach all 50i9 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

0 Applicability Review 50.59 Screening 50.59 Screening No. 6D-16-009 Rev. 0 50.59 Evaluation 50.59 Evaluation No. 6G-16-005 Rev. 0

Evaluation No. 6G-J 6-005 50.59 REVIEW COVERSHEET FORM LS-AA-]04-1001 Revision 4 Page 3 of 3 Station/Unit(s): Byron Unit 1/Unit 2 Activity/Document Number: OBOA PRI-7 Revision Number: 2 Title; OBOA PRI-7 Loss of Ultimate Heat Sink Unit 0 See LS-AA- I 04, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-16-007 50.59 REVIEW COVERSHEET FORM IS-AA-104-iOOl Revision 4 Page 1 of 6 Station/Unit(s): Byron Unit 2 Activity/Document Nunther: EC 406958 Revision Number:

Title:

flY Iso!ation Signa!s .DuringNormal Power Operation For Steam Generators W2BflC/2D NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of II) CFR 50.59(d)(2).

Description of Activity:

(Provide a bnef concise description of wbat the proposed activity invoive&)

This Temporary Configuration Change (TCC) Engineering Change (EC) 406958 will disable the Water Hammer Prevention System (WHPS) Steam Generator low level and Steam Generator low pressure feedwater isolation signals on Unit 2 after Feedwater (FW) 11 ow is transferred from the main feedwater isolation valve bypass valves, %FVVO43MB/C/D, to the main feedwater isolation valves, 2FWOO9AJB/C!D. (when power level is above approximately 25%).

The proposed change to disable the WHPS FW isolation involves lifting a lend from a terminal block for each of the

%FWOO9A/B/C/l) valves, installing an electric switchablejumper between two terminal block points for each of the

%FWO3%A/B/C!D valves, and repositioning the control switch from Auto to Open for each of the 2FWO35NBf(/D valves.

The WllPS FW isolation signals to the 2FWO43AIB/UD valves are not disabled because these valves are normally in the closed position above 25% power level.

The following changes will be made to plant operating procedures:

. %BGP 100-i, PLANT HEATUP, will be revised to verify that WHPS FW isolation has been enabled prior to starting any 1W pumps.

S %BGP 100-3, POWER ASCENSION, will be revised to disable WHPS flV isolation after the %FWOO9A/B/CID valves are open.

. 2BGP 100-4, POWER DESCENS1ON, will be revised to enable WHPS FW isolation for the 2FWO3SA/B/C[D and

%FWO3%MB/C/D valves prior to closing the %flVOO9AIB/OD valves, and then perform a follow-np action to enable WHPS FW isolation oo the 2flVOO9A/B/CID valves.

. 2BGP I 00-4T4, REACTOR Th.1P POST RESPONSE GUIDELINE, will be revised to verify that the

%flVOO9MB/C/D, %flVO4%NB!C/D, and 2FWO%9A/BICID valves are closed prior to starting the Startup Feedwater Pump. Action will also be taken to enable WHPS flV isolation later in the post trip response.

Reason for Aetivity:

(Discuss why the proposed activity is being performed.)

flns activity provides the capability to disable the WHPS Steam Generator low level and Steam Geuerator low pressure feedwater isolation signals during power operation above start-up levels to address an adverse vulnerability with the existing design. The existing design can cause spurious/unnecessary WHPS actuation, feedwater isolation, loss of normal feedwater flow, and reactor trip solely due to the loss of a single non-safety related power supply for the relays.

The changes to the plant operating procedures are made to administratively control when the WHPS flY isolation function is disabled and enabled.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases. or safety analyses described in the UFSAR.)

The safety function of the Feedwater isolation Valves is to close, isolating main feedwater flow upon a safety injection signal, feedwater isolation signal and high-2 steam generator level. As described in the Byron UFSA R, Several water hammer prevention features have been designed into the feedwater system. These features are provided to innnrnze the possibility of various water hammer phenomena in the steam generat:or preheater, steam generator main feedwater inlet piping, and the steam generator upper nozzle feedwater piping. Steam generator (two-out-of-three logic) low level trips are provided to close all feedwater isolation valves, feedwater isolation bypass valves and feedwater preheater bypass valves. Steam generator (two-out-

Evaluation No. 6G-16-007 50.59 REVIEW COVERSHEET FORM LSAA-iO4-1OO1 Revision 4 Page 2 of 6 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 406958 Revision Number:

Title:

Temporarily Defeat FW W;*teflammer PreventionSystem(WIWS) FW Isolation Signals During Nçrmal Powex OperatiopFor Steam Generators 2V%B/2C/2D of-three logic) low pressure trips are provided to close all feedwater isolation valves, feedwater isolation bypass valves, feedwater preheater bypass valves, and the feedwater bypass tempering valves.

The UFSAR described water hammer prevention features vere included in the NRCs basis for approval of (I) the designs of the Steam Generator and the FW/AflV systems, and the measures to address Unresolved Safety Issue (USI) A-I Steam ,

Generator Water Hammer; and (2) the removal of pipe restraints resulting from the elimination of arbitrary intermediate breaks in the FW system piping system design.

The purpose of the WHPS flV isolation is to prevent water hammer by precluding the admission of cool FW into a steam-voided preheater section ofthe steam generator. When the WHPS flV isolation is disahied at power, the safety related ESFAS signals function to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. A review of the postulated transients that could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 406958). The review determined that the ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip + Lo Tave) will function to prevent water hammer by precluding the admission of cool FW into a steam-voided preheater section of the steam generator during power operating conditions, Note that the %FW039 valves close on reactor trip and no coincident Tave is required. At low power levels or during startup the WHPS FW isolation features will be enabled. The design bases function to preclude the admission of cool FW into a steam-voided preheater section of the steam generator is not adversely impacted.

The other UFSAR described WHPS features will not be altered by this EC. Operating procedures associated with only providing feedwater to the upper nozzle of the steam generator during startup and low load conditions are unchanged.

Temperature monitoring and alarms to detect possible back leakage of steam from the steam generators into the feedwater piping are unchanged. Forward flushing and temperature permissive interlocks associated with opening the FWW are also unchanged. This tiC does not alter spht feedwater flow operation.

implementation ofEC 406958 does not impact the safety related circuits described in the UFSAR and Technical Specifications that generate an independent Feedwater Isolation Signal, including Safety injection (Si), SG Level Hi-% Trip (P44), and Reactor T.dp (P4) coincident with RCS Average Temperature Low at 564°F. Implementation ofthis EC does not impact the Reactor Trip signal generated on low-low steam generator water level.

The change will require new operator actions to appropriately enable/disable WHPS FW isolation In the plant startup, power ascension, power descension, and plant trip response procedures. Operators will reposition the control board hand switch for the 2flVO35AIB/C/D valves and reposition thejumper switch for 2flVO39A/B//D valves. Operations will also request Maintenance to either lift or re-land the leads for the 2FSOO9A/B/C/D valves.

. For startup WHPS FW isolation will be verified to be enabled prior to starting any FW pumps. Once the 2flVOO9 valves are open, action will be taken to disable WHPS FW isolation for the 2FW009, %flVO35, and 2FW039 valves.

. For a normal shutdown or low power operation action will be taken to re-enable WHPS FW isolation for the 2FW035 and 2flVO%9 valves prior to closing the 2FW009 valves, WHPS FW isolation for the 2FW009 valves will be enabled prior to defeating the ESFAS FW isolation signals.

. On a plant trip action will be taken to verify that the %FWOO9, 2FW039, and 2flVO43 valves are closed prior to starting the Startup FW pump. The 2FW035 valve flow path is used to direct flV to the upper nozzle. The UFSAR described main F?! flow arid temperature permissive interlocks will prevent inadvertent introduction of cold FW to the lower nozzle until action is taken to enable WHPS FW isolation.

There is no i mpact on any emergency operating procedures. The proposed operator actions are not time critical and do not impose a significant new burden on the operating crew.

Evaluation No. 6G-16-007 50 59 REVIEW COVERSHEET FORM LS AA 104 1001 RevisIon 4 Page 3 of 6 Station/Unit(s): Bxrowihdtl Activity/Document Number: ECAO6%58 Revision Number: Q

Title:

Temporarily Defeat Watçrflgmçr Prevention Syem ignais.Diiring QpcLndiQp çr Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

Above 25% power the safety related ESFAS signals function to close the feedwater isolation valves on a low steam line pressure safety injection (< 640 psig) or steam generator water level Iow4ow reactor trip S 36.3% NRj coincident with low Tave (less than 564° F). Note that the 2FW039A/B/Q/D valves close on reactor trip and no coincident Tave is required. A review of the postulated transients that could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 406958). The review determined that the ESFAS LW isolation signals (Steamline pressure ST or the Reactor trip + Lo Tave ) will function to prevent water hammer by precluding the admission of cool FW into a steam-voided preheater section of the steam generator during power operating conditions. Therefore disabling the WHPS low level. and low pressure F\?V isolation auto close signals above 25% power does not increase the probability of water hammer in the SG preheater and the main flY inlet piping. Thus the proposed change does not result in more than a minimal increase in the frequency of occurrence of an accident caused by a water hammer event in the FW line or the Steam Generator (flV line break or steam generator tube rupture).

The defeated WHPS FW isolation function is not used or credited to mitigate any design basis event, analysis, or Emergency Operating Procedure Operator Action. The credited automatic ftmcdon is the ESFAS flV Lsolation which is a completely separate circuit Thus the ESFAS FW Lsolation is not affected by this change.

The change adds new operator actions to disable and enable the WHPS low level and low pressure FW isolation auto close signals. Failure to enable the WHPS low level and low pressure auto close signals for startup, low load, or increasing load conditions could potentially result in a malfunction of equipment important to safety, specifically prevention of water hammer in the steam generator preheater and/or steam generator main feedwater (FW) inlet piping. The likelihood of occurrence of a malfunction associated with the new operator actions is not more than minimal based on the following:

. The actions to disable/enable the WHPS low level and low pressure FW isolation auto close signals has been added to the appropriate plant procedures and operators will be trained on the revised procedures.

S The actions to disable/enable the WHPS low level and low pressure FW isolation auto close signals do not have a required dine for completion. Above 25% power there is no specific time requirement for disabling the Wj45 low level and low pressure FW isolation auto close signals.. There is no impact on any emergency operating procedures.

The proposed operator actions for restart, power descension, and reactor trip are not time critical and do not impose a significant new burden on the operating crew.

I This change does not introduce credible errors during performance of the operator actions required for this change.

Three actions are required to block or enable WHPS following this design change. Lifting a single lead for each

%flVOO9 valve in 2PA27 or 2PA2SJ panel. These leads will he labeled as part oftids design change. The terminal board is clearly and accurately labeled. Manipulation of wires on a terminal board is considered a routine action with no credible failure risk. Lifting/landing the wire does not initiate a water hammer prevention actuation; it enables/blocks the circuit. The %FWO39A/B/QD valve water hammer prevention feature will be enabled or blocked with a switchable jumper. This jumper will. be installed across the original water hammer prevention actuation relay contacts, thereby allowing blocking or enabling the original contacts. This switch will he clearly labeled. The 2flV035A/B/C/D valve control switch in the Main Control Room bypasses the water hammer prevention circuit when it is placed in open (as opposed to auto). This is a routine method of configuration control by operators.

Lifting/landing ofthese leads for %flVOO9s, operahon of the blocking switch for the 2FW039s, and operation of the permanent switch for the 2flV035s will be controlled as part of a formal procedure. The wires to be lifted, switchable jumpers to be installed and used, and hand switches to be manipulated per this activity will he clearly flagged and labelled per the temporary Engineering Change to allow for clear and easy operator identification.

Maintenance procedure MA-AA-1070, which maintenance has trained on, exists to provide guidance for lifting and

Evaluation No. 6G-16-007 50.59 REVIEW COVERSHEET FORM LS-A.A-104-iOrn Revision 4 Page 4 of 6 Station/Unit(s): Byron Unit 2 Activity/Document Number: 1W %O695 Revision Number: Q

Title:

Tempprarily DefeatFWWater Hammer Prevenflion Svem (WIIPS)FW Isolation SignaisflurI ng .ormaI Power Operation qr Steam 9enerators2Ai%B/2C!2D landing leads and installingjurnpers and will be utilized in support ofthis activity. These actions do not initiate a water hammer prevention actuation; therefore, there would be time to recover from unlikely actions such as blocking or enabling WHPs out of the expected sequence.

S The system function to preclude the admission of cool EW into a steam-voided preheater section of the steam generator is not adversely impacted.

Above 25% power the ESFAS 1W isolation signals will function to prevent water hammer by precluding the admission of cool FMT into a steam-voided preheater section of the steam generator. Appropriate administrative controls will be put in place to enable the automatic WHPS 19W isolation function dating plant startup, power descension, and after a plant trip when the ESFAS FW isolation is not available. Thus the proposed change does not permanently substitute manual action for automatic action for pefforning a UFSkR-descdbed design function.

The WHPS low level and low pressure FW isolation auto close signals are classified as non-safety related. The WHPS UW isolation auto close signal for each individual generator is based on a two-out-of-three logic. The safety related ESFAS FW isolation signal on low steam line pressure is based on two-out-of-three logic in any one of the four steam lines, A reactor trip is actuated on two-out-of-four low-low water level signals occurring in any steam generator. A FW isolation signal is generated with the reactor trip coincident with a two-out-of-four Lo Tave signals. The safety related ESFAS PM isolation signals provide the same level of redundancy as the WHPS flY isolation signals. The safety related ESFAS components are considered to be snore reliable than the non-safety related WHPS components. The highest risk for water hammer events is during startup and low power operation when the ESFAS FW isolation signals may be bypassed and only the WHPS FW isolation signals would be available for water hammer protection. This same level of redundancy will be in place with only the ESFAS flV isolation signals providing protection when the VF1PS low level and low pressure auto close PW isolation signals are disabled above 25% power.

The proposed change involves lifting a lead from a terminal block for each of the 2FWOO9Am/tDl) valves, installing an electrical switchablejumper between two terminal block points for each of the 2FWO39MB/ciD valves, and repositioning the control switch from Auto to Open for each ofthe 2FWO3SA/BKDR) valves.

In accordance with procedure MA-AA-1070, immediately after the lead is lifted for the %FWOO9 valves electrical insulated tape will be applied on the exposed part of the connection, This will prevent impact to the associated 2FW009 valves or any of the other equipment within the panels. Thus the proposed change will not result in a more than minimal increase in the likelihood of occurrence of a malfunction of the 2FW009 wilves.

The switchable jumpers used to defeat water hammer protection on the %FW39AfBKID valves are built using a rocker switch and Pomona/banana jacks. These devices are built to handle much more current than the 3 Amp protection fuse for this circuit. The switchable jumpers are put in parallel to the WI-IPS contacts. In the event of a failure of a cable or couti ection, it would be a conservative failure that would leave WHPS in an enabled state. The Pomona cables have spring-loaded covers witich automatically cover the exposed male connection when the cable is removed from thejack. This will prevent the cable from shorting and closing the 2FW39VB/CID valves if the Pomona cable is inadvertently removed from the Pomona jack.

The switches used are similar to switches used throughout the plant and have similar reliability. Thus the proposed change does not change the likelihood of occurrence of a malfunction of the 2flV039 valves.

The control switches for the %FW035 vales are designed to allow the valves to be held open. Thus the proposed change does not change the likelihood of occurrence of a malfunction of the %FW035 valves.

The proposed change does not impact the other LTFSAR described WHPS features. When the WHPS fly isolation is disabled at power, the water hammer prevention flinction to preclude injection of cool water into a steam-vokled preheater section of the steam generator will continue to be performed by the safety related ESFAS signals to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. At low power levels or during startup the WHPS FW isolation features will be enabled. Additionally, the proposed change docs not rev ise the opurating proccdurc stcps tssociated with switclung teedwatir flow lrorn thc top fccdnozzlc to thc

Evaluation No. 6G-16-007 50.5% REViEW COVERSHEET FORM is-AAio4-1oOI Revision 4 Page 5 of 6 StntionlUnit(s): Byron Unit%

Activity/Document Number: ICC 4%9Sf 11evision Number: Q title limporarily Deftat FW Waterikmmer Preniibon Smtcm (WIIPS) FW Isolation Signals During Normal Power Qwfion .ForSteam.Cençra ors%A/%BIç lower main feed nozzle. As described in the response to PSAR Question 01 0+5 1 initial plant startup testing was performed to confirm that no damaging water hammer occurs when FW delivery is transferred from the top SG nozzle to the main feed nozzle. Since the transfer procedure steps are unchanged, there is no change in the frequency of a water hammer when FW delivery is transferred. Therefore the proposed change will not result in more than a m inim.ai increase in the likelihood of a water hammer induced malfunction. The results of the Stan-up testing are not impacted.

Disabling the WHPS low level and low pressure LW isolation auto close signals above 25% power does not introduce the possibility of a change in the coasequences of an accident because the FW isolation from the WHPS is not credited with mitigating any accident previously evaluated in the U PSAR. The ESFAS FW isolation signals (Steamline pressure Si or the Reactor trip + Lo Tave ) will continue to close the feedwater isolation valves as designed for accident mitigation. Thus the proposed activity will not result in a snore than .ndnimal increase in the consequences of an accident previously evaluated in the UFSAR.

Disabling the WHPS low level and low pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. because the flY Isolation from the WHPS is not credited to mitigate the consequences of malfunctions of an SSC important to safety. The proposed change does not impact the safety related ESMS FVY isolation circuitry. Thus there is no change to the UFSAR described failure modes and effects for the ESF actuation system. The proposed changes to the YFIPS FW isolation auto close signals does not introduce any new failure mode for the system.

Disabling the WHPS low level and low pressure flV isolation auto close signals above 25% power does not introduce the possibility of a new accident because the proposed change is not the initiator of any accident arid no new failure modes are introduced. When the WHPS FW isolation is disabled at power the safety related ESFAS signals will result in flY isolation actions that provide the water hammer prevention flincilon equivalent to the WHPS FW isolation. Administrative controls will be put in place to ensure that the WJ-WS FW isolation signals are enabled during startup and low power operation. The likelihood of operator error leading to an accident of a different.type is not credible because the actions are reflected in plant procedtues and operator training programs, the actions can be completed in the required time, the actions are simple and adequate time is available to recover from any credible error. The design bases function to preclude the admission of cool FW into a steam-voided preheater section of the steam generator is not adversely impacted, thus no new accident types are introduced.

Disabling the WHPS low level and low pressure FW isolation auto close signals above 25% power does not introduce the possibility of a malfunction with a different result because when the WHPS FW isolation is disabled at power the water hammer prevention function will be performed by the safety related ESFAS signals. Administrative controls will be put in place to ensure that the WHPs 1W isolation signals are enabled during startup and low power operation.

This temporary change to the Unit 2 WHPS Steam Generator low level and Steam Generator low pressure flV isolation auto close signals does not result in a change that would cause any system parnmeter to change. WHPS is not credited for mitigation in any accident analysis, thus there is no impact on any UFSAR described fission product harder limit. T..herefore, the proposed activity does not result in a Design Basis Limit for a Fission Product Barrier (DBLFPB) as described in the UFSAR being exceeded or altered.

The proposed change does not involve a method ofevaluadon as defined in.L.S-AA404-i000. Therefore, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in. the safety analysis.

Based on the results of the 50.5% Evaluation the activity may he implemented per plant procedures without obtaining a License Amendment.

Eyaluation No. 6G-16-007 50.59 REVIEW COVERSHEET FORM i.S-AA-1.o41oOi Revision 4 Page 6 of 6 StationfUnit(s): Byron Unit 2 Activity/Document Number: !c.41/2958 Revision Number: 2

Title:

Temporarily Jkfci.t FIV WaterJJmçftçveutign System (Will ) FWisolation Stenals Dunn g Normal Power c)ention For Steam Gen.erators %A/%B/2ç/%n Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

ci Applicability Review i: so.s9 Screening 50.59 Screening No. . Rev.

50.59 Evaluation 50.59 Evaination No. 6Q4fr007.. Rev. 0 . .. .

See LSAA4O4, Section 5, Documentation, for record retention requirements for this and all other 50.5% forms associated with the Activity.

Evaluation No. 6G-16-008 5:0.5.9 REflEW COVERSHLET FORM LS-A44G%-.OOi Page 1 of3 St.adoatnit(fl: Bwat StataUaI:ts I and.

ACthih?D*CUflhtntNUmhtr DSP I64S3TSB Ciangn 16411 Rt;tn Numbir: 0 TItk: Appendix 3 Scope Redadian ai a Rnsth of Trot N.rtk Engmnetrmn Report CommAssIoaed by Corporate Eagmadug NOTE For s&.sg .Eratnatons. mnfosotou on thLS fiçjfl protide tebasis forp7epath5thebieai umma y report

• ubmttted to tThe NC meccerac tnttba reopirernants of 10 CIR 5t594)(fl.

Dtsaipte: oIM:thity

[1roAde a brief. concise deccflptonofwbat the ppoied actvty imhe.:)

Theprtp:ostd acthityta documentaffonchange.addnssiitgtbeexdtrton from 10 CER SOApp:edxj Option 3.Type C local leak rate terhrtg (LU.Tj oftht folLowing Contairm.ent !soadon tAkes.

Penetration I Valve EPN I Dezcflpdo*n ChemicAl d Volume C:oflflI Syztem (CV]

1(2)CV&I5;2 AOVU-i(U-2) LTDWIIHDR ISOL nv (EUP VU]

P41 i(2)CV3160 AOVU4(U2) LTDWN ORIfiCES OUTLET HDR ISOLVLV (EOP VU]

1(2)fl8iOO )iOVt::.IQI-2) RCPPS:SEALL/O Him OUTMDECN?tT TSOL titr i.(2)CV8112 MGVV4QJ-2) RC P1S SEAL LjO HDR INSIDE CNMT ISOL P 3 lilt, 1.(2)CV8113 U-i(U4) RCP SEAL L/O HDRISOLVLtT i(2)CV8112 BliP HDRCHKflV ContaiatmenrSpray System (CS) 1(2)CSOO1.A IOY 1(2)AC5 PP DSCH HDRISOLVLV P 1 J2)CSOO.aA M2).A CS PP t(2)CSOiPASPL.Y.1OZL 14DB. CHKVLV P46 i(2)CSO07B: NOV 1(2)ECS PP DSCH HDRISOLVLV j(.)C.SOOBB: 1(2)B CS PP t(2)CSOLPB SPRWNOZL HDR CRKVLV tTSAR Secnou2.4.2.7 will Nnnaedto udeorthEC 4C4P72 a the &otirce docjuaith the exclaision from 10 CFR 50 Appendiz[ Option B Type C IozaI]eak nte tetrmz for thepeoertaiios aa valves above. UFSARTaMe 6243 wiilbe rerked to +/-thcn that Te C Sfl! not appIabLe far thesepaetrnou aM valves.

Technical Speciflcatoas Basee Table B3.63-1 wfl be* retsed to aoate that the penetrom and valves above are excluded fro*ni 10 CPR 50 AppeadixJ Option B,Type Clocal.leAk.rate testug.

The foUonigpmcedwes are irnpaced bythe exciusioa ttn 10 CFR SO Appendfrj Upton B, T:pe C)ocal ]eak rate testing:

I) BAP :600-lLkl 2 3\7 300-39 31 l*1BVSR..l1-24

4) lIBVSRó..i.l-25

5) l23.OSR&ll-P SI l2sosRatl-:4 Reaonfor Acth-IP:

Discucs why the proposed acthity m bang pci-foisted.)

The elimination oflocal leak rate terbng frr valves which meet the exclusion criteria wi!] reduce persornel ra&aton erposure and will decease work scope for refte]tng outages

Evaluation No. 6G-16-008 5O59 REVIEW COIIRSTIELT FORM Ls-AA-:c4-:oo]

Pst:ten 4 Paze 2 f 3 Statioatthtt): Byron Stathu/Cuib I and 2 ActhtIocuzneut Number: ThEY 164S3753 Changi 16411 RnkIon Nrnnber: 0 Tith: Appeudk J Scop Redichon is 2 Rnlt cf Tnt North Eadanrit Report CommAsioned by Corporate Engmeenng Effect ofActintv:

Discuss how the acnnrr urpacs plar op&atoas d:rn batas. or safem anaines eathed m the UFSAR)

This ctvtwrev:zez zke UFSAK Techmcad Spedficaon Baes and applicable procedures to reflect the eowhzion from

XFRSQ Appendix 1 Opton 3 Type C [OCSE lea rate tesng forthe contammett isolaton valves listed :n the Descnpton ofActnty se+/-on a5ove. EC 404972 and iupportng docnrnents uztfv th exclusion based on the current hcenzngbans rByron Station. The 3a52z forthis exdusion is given below.

Technical Speci&ator. .3.I6 Centa:nment Leak Rate Testing Program states that the program hd be n accordance

%9ththe gud&;nes contained tn RegulatoryGmde (RG1 1.163 Perfonnance-Bared ContatnmentLeak-TesnngProgram dated September 1995 with ecepdonz as noted in the TS. RG 1.16.3 waa isnied by the NRC to endorte the use of NEI 94-Ot.

In EI 94-01 Revision 0 Section 6 0 General Requirements. components that may be excluded from the LLRT requtrements of 10 CFRSO Appendix J are descnbed below:

Afl LIlT is a test performed on Type B and Type C components. An LLRT is rot required for the following cases:

S PriTna2y containment boundaries that do not constitute potentai primacy containment amosphedc pathways dunng and foUocing aflesIgn BasisAcodent (DEAl; S Boundaries sealed with a quai:Eed sea] zvstem or I Test connecton vents and drains bent,een prima containment isolaton valves whLch are one tnch or less tn tire adrninlstratvely secured closed and consist ofa double barrier EC 404972 demonstrates thatthe CV and CS valves in the De;cdflon of Acthr secdon above sats the atcena for exclason from Type C tesrz The valves will cnntnue to be tested in accordance with [a:eMce Testng Zfl Program requ;red by 10 CFR SOSSa. The affected vtes will also remain classified as containrnentizolarionvalves. No phytcal a ork is required to the ; alves or the system.

Summary of C onclusion for the Acthitv t SO.f.9 Review:

Pro tde juatficanon 1cr the toac hsson iodmthnr sufficient detad :o recormze md naientaod the e3senaai azrniens: lesthar to the conckmoa. Provide more than a uncle tncement that a 50. 1P Screenjir. 5G..9 :v:ua:o. or 3 Lsease Atnendnent Requtt. as applicable. s not required.)

A 30.59 Screentn g was performed which concluded that the acnnty require; a 3059 Evaluanon to implement the changes. The chanres to :ae documents map!emenugthe LRT scone reducoon unir EC 4O402 could be considered a

bange .o a prxethn RLRT p:oam nipienea.anon procedtres) tat may Ivemely affect how a IES.AR de:mbed de:gn frncthn is performed or coonofled. Because the i.cope ofTvpe C fliT testing s bseng reduced. the effec: ofshe proposad

banse : conserranve]v cons:dered to he adverse. for the pucposes of evalwnos

.4 5OS9 Evaluation was performed whtch concluded thatthat the documentation changez can be imp)emented per plant procedures wthout obtaining a bcenzeAmendment Theprcposed chaages to the UTSAR. Technical Specicanon Baset.

sod aftcable procethwe do not tequare a :hange to Techmca2 Specs&anoas becaure Jy do oct al:et or mo&f any ensttzg

ercaa Specth:anon. The basis for this conclusion to presented below.

The prcpote acnnn ofren:m! :be LTSAt Techwcai Specthcanon Bates. and appbcable procedue do not phyccallv alter ny eqiupinea:. 5e5Z p&tOLfli&e. or opera:or acflosa that cotild aect the acafeot ni Dans:ent: rn Chap:ess 5 and ;; of the LTSA.R or their frequeocy of occureoce Therefore. the ioctmien anon ckaavt related to anplemenung the LflT stope reducnon under EC 4(4972 do not sonthice the pocstbilcy of a change in the frequency of occneace of an accideoc prenonaly evolua:eo in the tTSJY.tecauie the lethare .ofthese valves ts not ann uzna:o: ofan accdenc andzo new fatwe modac are mroduced.

Evaluation No. 6G-16-008 5059 REI1EW COVERSHEET FORM Ls-AA-:c4-:oal Revttou 4 PagE 3 of 3 Sbboatthrs) Byron StaáemZUthts I and 2 Acthirv:Decuxnent Number DRY I64SVTSB ( bant 1641] RevLwn Numbir: C

Title:

Appendix I Scope Reduction is a Renilt ofTrue North EnzmueerinE Repon C ommn:ioned by Corporate Engineehug The ::opoted dccnaenra2oo chaogt do not phycaily aEer an eqwpneoz. snten pedore. or opentor ac20ns that could afftct me pornae of the dettgn ftmctca of :he conttmrn: itolartoa vthe associacci wn thete chanres TAo :tnenr tTSAR anaivse renam boimdlar Therefore. the aoctneaancu changes nplemented nuder this actww do o: reulit n nore than znnuza] irese m :ae hkelthood ofocazreace of a naifncaou of n SSC anperan to afem pretously evatma:ed in the L7SAR.

The rcpoted actvn does oot retith u auc áanre3 o the operation ofhe affeaed ss:en CV and C S: :hat could impact the uitersim d :edaancy of those nntm in a manner that could :ethice the ability to perforni their inreuded dea fancnous Therefore the proposed acthi doe uo: re:tilt n note than a rnrniznil macate in the conreoyecet ofan accident p:etoutiy eaula:& in the LTSAR The affected CV arid CS peuetacons nd vaIve re atduded from Type C URT on the basis thatthev do not constitute a potenta] pnnuiy contairsuent annowbedc path nay during a desgn basis accident therefore this acmitv does not renltin an increaze frithe courequenres ofa nnthnrton of an SSC hzportantto safety previously evakiatedin the UFSAR The documentation changes uzpemanted under thit 3CUfl7 do not phyticallv aher any equipment syttem perforniance. nr opentor actions that could affect the pethrnnnce of the deagn nuiczon of the containment isolanon ahes asiociated ith theSe chansec. The current UflAR analnes reman bonding. Therefore. the proposed xtvzry doet not ceate a poibthtv for an accident of a tiferen: type than nix previously ia1uated rn the ursn Renionig Type t LLeilJ teatng requirement; for the affected peneatons and containment inolaton valves in accordance wfth approvedreguiator; docunentation does not irnodure the possibthtfor a maThinrüon ofan SSC with a different result becaure the acdviw does not intoduce any new failure no des or failure results. Therefore the proposed acthft, does not ceate a posnbthwfora itaifuncdon ofan SSC icporttnt to safew with a thfferent rerult than any prenouLly evaluated in the UPSAR T;pe C LLRT ü being removed on the hasis that the valve cannot provde a leakage p adwav for corxamrnent abzosphen and therefore ensures that La the maxImum allowable leakage rate s not :mpacted This actvty does not result in a rhange that would cause any nysrem parameter to change Therefore the ac&ttv does not resulr in a fission product barner (and; or a DBLTFEJ as deacifoed in the UFSAR be:ng exceeded or altered The thanges to the LFI4R Technical Spectflcation Bases and procedures re)ated to inplensening the LLRT scope reducnoa under SC 4)4972 for the affected CV and (S peoeUauoat and valvet do not :n-olve a method of raftanon as described to the UFSAR La s not being changed as a result ofthts aah-i Therefore the acthttv does not renilt in a departure from a method ofevaluahon. This actvir; does not :mpact revise or replace the methods used to deternune pnniaiy contatoment eahage

Attachment:

Attach all 50 ;g Review fonna completed. at appropriate.

Forms Attached: (aeck all that apply)

C App&abelity Review o9 Screening Q EQ Sceeniag No. 6E-1&4t9 Rev. U G O.S9 Enkadon f0 59 Li aluaton No. 6G-WDOS Rev. 0 ee ZS-t-lD4. tecnon 5. Docurnennoa. Sir record recennon requrernents for tha and all otter 50.5P farnis attacated with the Actvn

Evaluation No. 6G-17-OO1 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of2 Station/Unit(s): Byron I UnitOj Activity/Document Number: EC 618443

• Revision Number: 0 Tide: Lost SEP Crimps and RVLIS Pins ..

NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 5059(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

During the performance of an annual spent fuel pool (SfP) lighting inspection, Reactor Services discovered that five lighting lanyard safety clip crimps are missing (JR 39755 1 1 ). The crimps are made of 304 stainless steel with dimensions of approximately 0.375xO137xO.229 (Qty. 5) oval crimp fittings. Reactor Services performed an underwater search, but was unsuccessful in locating the missing parts. The activity is to evaluate the effects of this potential foreign materIal in the Reactor Coolant System (RCS) and Spent fuel Pool (SFP). Additionally, two locking pins were found missing from RVLIS connectors. These are 0.170 long x 0.1 diameter shaft with a 0. 125 diameter head and are made of stainless steel. These were lost in the refuel cavity area, Reason for Activity:

(Discuss why the proposed activity is being performed.)

Loose parts (5 stainless steel lanyard crimps) are missing and potentially dropped into the Spent Fuel Pool (SFP). Based on IR 39755 1 1, these lost crimps were located in the northeast corner ofthe spent fuel pool over discharged fuel. These could not be located and are potential foreign material that could have fallen to the bottom of the SfP, or could be on or in a fuel assembly. A likelihood exists for some or all of the crimps to be transferred to the Unit I core during B 1R21 ftiel reload.

Additionally. two very small RVLIS locking pins were lost in the refuel cavity area (IRs 3979800 and 3981072). These were located at the drawbridge from the refuel cavity to the reactor head. A lost parts evaluation is required per ER-AA 2006 to evaluate the potential consequences that the missing crimps and pins may have on plant equipment.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UfSAR.)

There is no effect on the operation of any system. The location of the five small crimp fittings is not known other than they came from light lanyards along the northeast corner of the spent fuel pool over discharged friel. There is a small potential that these crimps could be transported into the Unit I RCS during core reload. The location of the RVLIS pins is not known, therefore, it is assumed for this evaluation that the pins could reach the RCS.

The potential for damage to plant components is considered less than minimal based on their very small size. If they become lodged in a fuel assembly, there is the potential to cause fretting to the fuel rod cladding. However, the likelihood of fuel clad fretting and failure is low based on the geometry of the crimps and pins (they are not long, stringy or spiral in shape) and the limited number of crimps (five) and pins (two) that might find their way to the core.

Summary of Conclusion for the Activitys 50.59 Review:

(Providejustification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

ECs 618443, 618441 and 618648 evaluated the effects of the foreign material on RCS, RCCS and fuel components. The loose parts will not increase the frequency of occurrence of a stuck control rod, RCP locked rotor, RCP seal failure, RCS decreased flow, or cause a steam generator tube rupture. Since the likelihood of a accident is not increased and the components needed to mitigate the consequences to an accident are not more than minimally impacted, the proposed activity does not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.

If the crimps and pins are transported to the RCS, the potential to damage various applicable components within the RCS and interconnecting systems was evaluated. It was concluded the foreign material. does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The introduction of the crimps and pins to the reactor may cause fretting of the fuel cladding, however UfSAR section 4.2 states dose limits given in IOCFRI 00 and IOCFR5O.67 will not be exceeded due to a small number of rod failures. The likelihood

Evaluation No. 6G-17-OO1 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 2 of 2 Station/Unit(s): Byron/Unit OJ Activity/Document Number: EC 618443 Revision Number: 0 Tide: Lost SFPrirnps and RVLIS Pins .

of multiple fuel rod failures is unlikely. Therefore the fuel, vessel, RCS. and connected systems and components would not malfunction in a manner different than has been previously evaluated in the UfSAR. Thus the consequences of a malfunction are not changed.

The loose parts are very small and would have minimal impact. The parts do not provide a means to initiate an accident of a different type than any previously evaluated in the UFSAR. The components have been evaluated in ECs 618441, 618443 and 618648 and it has been determined that they would continue to perform their design functions and any operational disturbance would be within the plants design and operational basis. Thus the loose parts would not result in an SSC malfunctioning with a different result than what was previously evaluated in the UFSAR.

The three fission product barriers (fuel clad, RCS, containment) were considered. A fuel failure in itself will not exceed or alter a design basis limit for a fission product barrier (DBLFPB) a fuel failure in itself will not exceed or alter a design basis limit for a fission product barrier (DBLFPB) because per UfSAR section 42, dose limits given in IOCFRIOO and IOCFR5O.67 will not be exceeded due to a rod failure. Therefore, the loose parts do not result in a DBLFPB as described in the UFSAR being exceeded or altered.

The loose parts evaluation does not involve any evaluation methodologies used to establish design bases or safety analyses.

It is concluded the loose parts will not cause more than a minimal impaci on plant components and systems.

Based on this evaluation, the proposed activity can be performed without prior NRC permission per the applicable governing procedures.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

forms Attached: (Check all that apply.)

EJ Applicability Review 0 50.59 Screening 5O59 Screening No.

• Rev.

50.59 Evaluation 50.59 Evaluation No. 6G47OO1 Rev. I See LS-AA-l04, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-17-002 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of 4 Station/Unit(s): Byron Units 1 and 2 Activity/Document Number: EC 400957 I EC 402963 I DRP 17-015/ DRP 17-016 Revision Number: 000/ 000 /N/A/N/A

Title:

Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-i Outage)

NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of iO CfR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description ofwhat the proposed activity involves.)

The proposed activity (EC 400957 for Unit 1 and EC 402963 for Unit 2) replaces components within the rod control logic cabinets (1RDO7J and 2RDO7J) as part ofan overall phased project to upgrade key non-safety-related process control systems, integrating them into an Ovation-based distributed control system. The upgrade is being performed in accordance with the recommendations ofthe original equipment manufacturer (Westinghouse). Separate activities (EC 400959 and EC 402965) are installing the Ovation distributed control system infrastructure (hardware and software) necessary to support the upgrade of the rod control logic cabinet. Separate activities are also installing an upgrade ofthe Ovation-based digital electro-hydraulic (DEH) turbine control system and connecting the upgraded DEH system to the DCS (EC 40095$ and EC 402964). However, the turbine DEH control system will be connected to the DCS at the same time the upgrade ofthe logic cabinet is connected to the DCS.

This activity will replace the existing logic cabinet internal hardware and components. The majority ofthe new components will be furnished as a pre-fabricated panel assembly sized according to the dimensions ofthe existing logic cabinet, which will be screwed to the existing rails inside the cabinet. New components will consist ofredundant controllers, input/output (110) modules, media converters, power supplies, line filters, circuit breakers, surge suppressors, a power distribution module, a transition panel which allows connecting to additional branches ofIJO modules, a temperature and humidity transmitter, and terminal blocks.

The upgraded rod control logic cabinet is being designed by Westinghouse, the original designer and original equipment manufacturer, to reproduce the existing logic cabinet functions using new hardware and software. The rod speed and direction signals will be reproduced using new hardware and software. Instead of being adjustable via potentiometers at the cabinet, these speeds will be adjustable via the software server I engineering workstation. In auto mode, the rod speeds will be developed from the analog input using new hardware and software in the logic cabinet. Existing rod speeds and stop interlocks will be maintained.

The potential for a single failure causing the rod speed to reach 77 steps/mm will be eliminated, such that the maximum speed will be limited to 72 steps/mm. The existing bank sequencing, bank overlap, and group alignment functions and the existing timing profile will be reproduced using new hardware and software. The bank overlap and timing profile will be adjustable via the software server I engineering workstation. As at present, overlap between successive banks will be adjustable between 0 and 50%

of bank height. The upgraded rod control logic cabinet will rely on redundant controllers to perform functions that are at present performed by various cards. These functions include development of the timing profiles sent to the power cabinets, group sequencing, and bank overlap.

New internal cabinet cables will be installed that connect the Ovation I/O, bridge rectifier circuits and relays to the existing field terminal blocks located on the sides ofthe cabinet. The pulse-to-analog (P/A) converters in the DC hold cabinets (1RDO$J and 2RDO$J) will be removed and replaced with analog output signals originating in the logic cabinet.

To provide maximum fault tolerance for loss ofpower conditions, the upgraded cabinet will utilize three independent AC-power feeds. The existing power feeds provided to the cabinet from the motor-generator (MG) sets will be retained by this modification.

A new third power feed will be provided from a spare circuit breaker in existing 120/240 distribution panels 1VE13E and 2VE13E. A new uninterruptible power supply will be installed in cabinets 1RDO$J and 2RDO$J to enhance the reliability of this power source. Auctioneering ofthe power sources will be provided by a new transfer switch, also installed in cabinets 1RDO$J and 2RDO$J. In the event that one ofthese supplies fails, the other power source will continue to provide power to the Ovation system, assuring normal operation of the system.

In the control room, the existing 100 VDC electromechanical rod group step counters, fifteen per unit, will be replaced with new 24 VDC digital step counters ofthe same form, fit, and function. The rod control step counters will be mounted in the existing step counter housings in panels 1PMO5J and 2PMO5J. The existing rod in and rod out lamps will also be replaced with new LEDs/lamps that are compatible with the 24 VDC circuit. Since the portion ofthe circuit that is being modified from 100 VDC to 24 VDC is also used by the loose parts monitoring system (LPMS), the LPMS circuitry will be modified to accommodate this change.

Evaluation No. 6G-1 7-002 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page2of4 Station/Unit(s): Byron Units 1 and 2 Activity/Document Number: EC 400957 I EC 402963 I DRP 17-015/ DRP 17-016 Revision Number: 000/ 000 iN/A/N/A

Title:

Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-i Outage)

The rod bank position output signals to plant process computer (PPC) cabinet 1CXO2J and 2CXO2J will be abandoned. This data will now be provided to the PPC instead via the Modbus interface from Ovation to the PPC, installed as part of a separate activity (EC 400959 and EC 402965). The rod supervision PPC software will be revised by this activity to support rod bank position data being delivered to it via the Modbus interface to Ovation rather than the existing method of insert/withdraw pulses being read by the PPC I/O system. In addition, rod bank position information for group 1 as well as group 2 rod positions will be used in the program, whereas the existing program only uses group 1 data.

DRP i7-015 (Unit 1) and DRP 17-016 (Unit 2) will revise UFSAR Section 7.7.1.2.2 to incorporate changes associated with the Ovation-based control system upgrade.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

The upgrade ofthe rod control logic cabinet is one ofseveral upgrades ofnon-safety-related process control systems being performed because the existing 7300-series control systems have exhibited performance, maintenance and reliability issues. It has been determined that the appropriate long term solution is to upgrade the existing 7300-series equipment to modem digital controls to enhance its reliability and overall lifespan.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

The basic operation ofthe rod control system will not be changed. The upgrade ofthe logic cabinet will involve minor changes to the operator interface with the rod control system (e.g., slight differences in the appearance of the step counter displays and in-out lights).

The functional and performance characteristics of the existing rod control logic cabinet will be maintained. Therefore, the proposed activity does not impact the design bases or safety analyses described in the UPSAR.

Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The proposed activity involves the replacement of components within the rod control logic cabinets (iRDO7J and 2RDO7J) as part ofan overall phased project to upgrade key non-safety-related process control systems, integrating them into an Ovation-based distributed control system. The upgrade is being performed in accordance with the recommendations ofthe original equipment manufacturer (Westinghouse) to address equipment reliability and obsolescence issues. The upgrade was designed to match the existing functional and performance requirements ofthe rod control system.

Evaluations ofthe Ovation platform and its successful use in non-safety-related process control applications in the nuclear industry indicate that the system is expected to perform dependably. The application-specific software is being provided by Westinghouse the original equipment supplier, who is employing their established product development processes used for the nuclear industry to match the existing functional and performance characteristics of the rod control logic cabinet. The basic operation ofthe rod control system will not be changed, although the upgrade will involve minor changes to the operator interface with the rod control system. Therefore, the operator interface with the upgraded rod control system does not fundamentally change or adversely affect the methods used by the operator in performing or controlling a UFSAR-described design function.

However, the upgraded rod control logic cabinet combines functions previously performed by multiple cards into redundant controllers, introducing the potential for failures to create new malfunctions which could adversely affect a UFSAR described design function. In addition, connecting the upgraded logic cabinet to a network linked to other plant control system introduces the potential for a software-based common cause failure that could adversely affect multiple systems. Therefore, 50.59 Evaluation 6G-17-002 was performed to address these issues.

Evaluation No. 6G-1 7-002 50.59 REVIEW COVERSHEET FORM L$-AA-104-1001 Revision 4 Page3of4 Station/Unit(s): Byron Units 1 and 2 Activity/Document Number: EC 400957 I EC 402963 I DRP 17-015/ DRP 17-016 Revision Number: 000/ 000 IN/A/N/A

Title:

Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-i Outage)

The 50.59 Evaluation determined that the redundant hardware and improved diagnostics combined with the reliability ofthe new equipment and software and the fact that single failures remained bounded by existing single failures (as determined in a failure modes and effects analysis) did not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UF$AR or in a more than minimal increase in the likelihood ofoccurrence ofa malfunction ofan S$C important to safety previously evaluated in the UFSAR.

The 50.59 Evaluation also found that an analysis ofthe susceptibility ofboth the Ovation platform and the DCS application-specific design features to potential digital system failures including software-based common cause failures was performed using the guidance ofEPRI Technical Report 3002005326 (Methods for Assuring Safety and Dependability when Applying Digital Instrumentation and Control Systems). In accordance with the EPRI guidance, the analysis considered the following potential sources of failure: random hardware failure, environmental disturbances, design defects, and operations or maintenance errors. Since individual device hardware failures had been previously addressed in a failure modes and effects analysis, the susceptibility analysis addressed hardware failures in shared resources. The analysis of design defects as a potential source of failure included a consideration of software-based common cause failures. The susceptibility analysis reviewed the design with respect to the various EPRI-identified preventive and limiting measures to determine whether these measure were sufficient to reduce the likelihood of a common cause failure to Level 2, which is defined in the EPRI guidance as a failure likelihood at or below that of failures considered sufficiently unlikely that they would not typically be postulated and analyzed as part of a plant safety analysis report (e.g., random hardware failures or common cause failures of identical components caused by wear out and aging or by design or maintenance errors). The susceptibility analysis found that the design was adequate to reduce the likelihood of common cause failures in the Ovation platform or the specific DCS application to the desired level (i.e., EPRI Level 2).

On the basis ofthe susceptibility analysis, the 50.59 Evaluation concluded that the increase in the likelihood ofaccidents or malfunctions due to a software common cause failure is no more than minimal, and that the activity does not create the possibility of an accident or malfunction of a different type. The radiological consequences of previously evaluated accidents or malfunctions were not affected.

The same system parameters will continue to be monitored and alarmed with the rod control system upgrade, and the same decisions and actions will occur in the control ofplant equipment and system during transients. The changes to the logic cabinet do not affect the overall operator response time and do not increase operator burden. The upgrade does not introduce any changes from manual to automatic initiation (or vice-versa) of functions. The upgrading of operator workstations does not fundamentally change the data presentation or the operator interface. As a result, the method by which the operators perform or control the logic cabinet design functions is not adversely affected. Thus the proposed activity does not involve a change to any procedure that affects how SSC design functions are performed or controlled.

The upgrade ofthe rod control logic cabinet in accordance with OEM recommendations does not involve a method of evaluation described, outlined, or summarized in the UFSAR that is used in the safety analyses or used to establish the design bases.

Supporting analyses of electrical loading, structural loading, loop scaling and setpoints, and HVAC loading changes do not involve changes to any UFSAR-described methodology.

The proposed activity does not involve a test or experiment not described in the UPSAR. The upgrade ofthe logic cabinet system does not affect the Technical Specifications or the facility Operating License.

Therefore, the proposed activity can be implemented per plant procedures without obtaining a license amendment.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Evaluation No. 6G-1 7-002 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page4of4 Station/Unit(s): Byron Units 1 and 2 Activity/Document Number: EC 400957 I EC 402963 I DRP 17-015/ DRP 17-016 Revision Number: 000/ 000 IN/A/N/A

Title:

Westinghouse Ovation Digital Upgrade for Rod Control Logic Cabinet (N-i Outage)

Forms Attached: (Check all that apply.)

0 Applicability Review 50.59 Screening 50.59 Screening No. 6E-17-024 Rev. 0 50.59 Evaluation 50.59 Evaluation No. 6G-17-002 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-17-003 50%59 R.EVIE\V COVERSIIEEl FORiI Pviion 4 Pi:.e t I St;tiorr1. r;ii(s): fl\H1fl1ittl&2

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Evaluation No. 6G-17-003 50.59 j)j\t{[4/ (X)VLRSJIEET Fi)R1 Revision 1

!JC 1 43 S1;llhin/i..iiit(s): [h tn LnU I&2

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[)Lf I s stein. ;s; a rLsuIt. die nudhhicd C)va!l pIt1brsn .tnd Uc rak.d I)h I ccnhrohs arc cxpcct:ci to pcrIOLI{i depcrd tbb I 1o cv:1% C u1;iecL nz the 1)1.: I Nstc:n1 I ci a ncttk h nik cd to oth:i itnt conrch s. s ens introduces t hC pUlciUiflI it)I a ixisee etiineu CâKC 1iI I tFC Uiit could aLli crsclv thhict iiuhiphe t:iii this \US c(Nisi(lclccl a dHz1;c to 1i1 sSC that di.ci <;C 3 at tects a I.. I S\R csc-tlcd dcsig,n hinctk:i; thcrc tcrc. a 5(35) c,utttaiinn b(J I 7 Oh)3 s

• ihe 5fl59 [(1htL1t1L)fl iOUfld Unit an anal, sis oh the susccpt bihit> oh t)Oth the t)vatinn phuttonu and be [)CS :ipphic;itionspcc Bc decn h;.aturcs to 1:Lcttt:iI dichtal , steni ,ti hircs ii. hti(lHk sti:ttvare4asej coinniun cuic EiI:tcs \\ as ;rfl nd u%ng the gudance ul t:ti I h echn cii [kcptt (1t)2OO532% (N*1ethiQds for ,\ssurin% Sutist a1 i)cpcndabihit.s ii hen AppI3hnsz t)it;it;t!

[fltrltIiftfl!tItHifl ailt (tutco1 Sstetn, ). In iccnliiance itli the EiR I nidince. tie n,hy:NIs consk{crc: the btbt,ine prcnhitI

.uLrc.s or fl.i iure: randotn  ; ttt hiia. en v ircUit:cntti t &IiSUh1tt1LCCS. tleZ.iLUi Jefcts. an.] itc*ttI1 ions ci nüutcii:tncc crrur.

Since IIJR!dcai devcc ht di:c a hiBjuics had hecti eautN!V addressed in a fiturc noecs nd ct[ccts analysis. he suscepUbuilt, aiah, sis addresscd bird i are au lures ;n shuircd resources the a: cdx sus of desien cclccs as a potcntLd stu cc oh ire hi: {UL12d a (3Iisiucratuon Of ,,ohrvar:4asecI eoiimnii cttisc iuIires. lie st1.5cepBauti:y unahy sis res iccd the tk!sien i ith espect o the S iN I t nfl led t- t lHtt id h u ri i tL tLLL1 U t ihchtt th,,sc. ut Stilt S tt_ il 1 CR it N !t hit h ukC:u;1(t of a ciiinci cause thilurc to 1 :vel 2. I c:el 2 is de JncU in he EPRI guiJinc. as 1iu hikc!uhni at or hLhns that

üt h:iulurcs c()5udcrCd su thctcit1 tint kels titer rhc ctd ot typrcahly he f)ostuateti and anahy.cd as b15rt oh t i:int ,ctch1 ZiitiSIS ror1 (ct. ; I1Ctt)1fl ia: ct are 1n ures or cuni n cause tUb eves o r dent ical cotupencu Ls cac ed b s ci ciut r1d tjai .

or h> d:sin or niamtcnancc crtt;rc). t h sstsceptuhit afl:ti? sus lound tI;at itie JcBt!n ias :idecpatc to rednea te ukehuhohi ekifihimiti cause thitacs in the (.) ciliUH platform or the specutic ocs Icitu;i to the desired 1e ci (i C. LPRI I eei 1).

()o the tsis of the susceptuhutitv :uiibyis. the 5059 [ vauathon corchided that the uricrese hi the iukct]iooJ o I accidents or lItItIUtlCUOflS thie tO 4 soinvare coiunon catso ut1Al C 15 [10 IlIore than niunniaI. .uicl thu ttiL atus ut. does O)t C1tctc Uc flc.suhht:

otiti acci,tczt or mat tI.it. tion ofa chtfercnt is )c the tbht. catscqucaccs of prevuoi.uy eratu:ited accidctcs cr cflattoc1ctuOtt scre not tttcted, 11cc vicicr chaniics the orerator ictcrhace vith the uptr;cdcd 1)1] 1 y,teni do ci.e uris dse a clI,tci.rc Li U ac ocodure thot a,h cc sch3 atYcts ho t : 1 S tkdcsccihc I SSC he I en tcc!lctcons arc pcrtriort or cntccB cd. t he i:e ocatec 1:O hjt ita cichic cc ;ot;cre Is sIc lar to c. st in e. oct h no cntack :uttccs. o, h cli ittcos tuc a ripid eJuctccn I a tcub;ne icd c cacttr p{n or to rcvent Ii1 11SCt cocccItIofl hunt hecLtIng to a nDcctot trip. \sh;IIc recciaic,inu. setI vitIcicc the iii tjccBorc cIirXIIutV arctIc CactC)r CQt1Uc)1 jfld stetni accap ceatrot S35tC1115. rIlerctC!a. this Os])eCt ut chic tro;scd activit, does flOt c1varcI% autct a U rs\Rdescruhicd [cozn Itic_tion I hisS ii ft ,fi 1 It it t st ILS 1 50 1 1 1 U 1) d coo I ci \It 12 Lt 1 b\ Dc O\ idttte tci.J hc tfln tI c rejircgranirtied r;nurc rate and pos or level t) Iocnd the ;;scr reciuctccn reqe it a I ucce the cacad ag, efltt at heater sit i I Icic SOiJtl&it ci e :aken into account. [he ness rjnT.ac: tiatue vuii c:ticc a ta;tnuai action hut II rD,kcce the operator taurchelL \ th rio cciv retertici tiihttrc cnodcs in cite 1citettctIfl of (;flcr:tto; S 55 uhc the piant s stenis. hcretore. iti accora_tree si ttti cite rucd*tctcc rrc ideri In Section 52.22 oI iS-:\ A I O:1_ I t:h1 ( StJ.59 Resource 1atictriit this acjiect at the proosed aCtIiIt% GOeS not cc; oRe a cti:tiec ho a prcccttre that aci crsciy ttfIccts d;: fltCthiti{.l tO ocitortu rg or control inc a t BSA1&tesccIhed dusl:tc iiccctcn.

[he upgrade oi the L)E[{ 1JCS lo accordance ith ( )t3i econicuenciatlins and the intrcxiuctuon oh the n\ n.cahack tcturc tk nct u VOIVC a ilett]G(I of c ;tucLji)n dcscrzoed. (clith nod. or sctt:iniartied in the LFS\R that Is used hc the c1tv ,cniics or used to

%) it c h Bk JL\t r 1 L -ducttonstis rc ci.. O , s c t _Lt t L O, I n nfl IL It, iLitti. I i\ V

• I ,

i acid ccuiarnect cisrnic qntit icatcon do not hnoke toy changes o I F\R descriied niethcdohopas. risc proposed actis tv iucs not iive a test n c.1scricc]cctt not dScrlDed ti the U IS:\ i? I ice ccp.rade otfite 1)iSi I Nystetn Lit)O5 not lttcct toc tccnccicai Sped uicatfons or the j::c ( )pcl titioc t icctcse.

I ihnciorD. NR(. icpp;ic oh I, not cecpcired tr the ripd :cBiis . lice aCtRtts tt\ iC tntpicnrertod Pot :h apiic.chi gcncnicng rn c cdure.

Evaluation No. 6G-1 7-003 5O5% RIWJ EW COVE:RS ILEET FORM RCViS14)fl 4 3o 3 Sntionftuit(s): Uui:%nU1llL1_.1 Acthity/Doeument Number: rC 4OO9fS 140%96% 1{evisiou Number: OOtt(fl)O FitIe Westinghouse_Qa1jiiDiga1UpgidijjJ-jQjgç)

Attach ments:

Attach all 3Q39 Riew ItHifla a)!flphC[Cd. as apprt)prfli!C.

fOrfliS Attachcd: (Check all that apply.)

ci Applicability Review 0 5O5% Screening $059 Screening )* (hl 7O%5 Rev. (1 r

S(L?% kiakuihion 5O5% [:nli.atiii No 60- 1 %M03 Rev 0 Lii Sec LSA1\I 01. Seciion 5. l)ocumentation, or record retention rcquIreiucnts tot this ani al other 54L%9 bhrrns assocuited uth the Activity

Evaluation No. 6G-1 7-004 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of2 Station/Unit(s): Byron I Unit 1 Activity/Document Number: EC # 619975 Revision Number: 0

Title:

TCCP to Install Pneumatic Jumper to Open 1WG074 NOTE: for 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief concise description of what the proposed activity involves.)

Temporary Configuration Change Package (TCCP) EC 619975 allows the temporary installation of a pneumatic jumper to allow opening of 1WG074 despite the associated solenoid valve, which is intended to automatically close 1WG074 when the Feedwater Pump Seal Water Collection Tank level is low, being de-energized due to a loss ofthe associated control power.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

IR 4018759 documents that MCC 133Z2 unexpectedly lost power due to the MCC feed breaker (1API8E-3C) tripping open. As a result ofthis MCC being lost, a number ofcomponents failed to the de-energized state. A solenoid valve associated with 1WG074, which is normally powered from MCC 133Z2, causes 1WG074 to fail closed when the solenoid is de-energized.

Closure of 1WG074 isolates the outlet from the Feedwater Pump Seal Water Collection Tank (1WGO2T).

JR 40 1 8774 documents that a pneumatic jumper was installed as an immediate action to place the situation in a safe and stable condition per PI-AA-120 to allow opening of 1WG074 to restore flow to the Unit 1 Condenser, which stops overflow from the Unit 1 Feedwater Pump Seal Water Collection Tank to the Turbine Building floor and thereby conserves Condensate Storage Tank water. The installation ofthisjumper is a TCC and requires documentation per CC-AA-112.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

This activity restores the ability of 1WG074 to open and allow level control ofthe Unit 1 Feedwater Pump Seal Water Collection Tank by bypassing the associated solenoid valve that automatically causes 1WG074 to close upon a Low Level condition in the tank. This activity modifies the non-safety related control loop to allow continued operation despite the loss of power to one component within the control loop. The level control loop normally functions to automatically control tank level using the output of a level controller to proportionally modulate the valve position of 1 WG074. The level control loop also receives a signal from a redundant level switch, which functions to close 1WG074 ifthe redundant level switch senses a low level condition due to a malfunction of the normal level controller. This TCCP will bypass the level control loop and tank level will be manually controlled by an operator continuously stationed at the tank. The automatic isolation upon a low level alarm from the redundant level switch for the associated tank is also defeated as a result ofthis TCCP.

Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The UFSAR does not describe or credit the Feedwater Pump Seal Water Collection Tank Outlet Valve (1WG074) or the associated Feedwater Pump Seal Water Collection Tank level control loop or its function. The automatic isolation function prevents an air inleakage path to the condenser from being established in the event that the tank is completely emptied. Loss of condenser vacuum and turbine trip are listed as initiating events for decrease in heat removal by the secondary system transients in UPSAR Section 15.2. Per UFSAR 15.2.3, this is a ANS Condition II transient ofmoderate frequency. However, the condenser has design features to maintain vacuum such as the SJAEs and vacuum hogging pumps. Existing station procedures direct operators to take action to maintain condenser vacuum. In the event of inleakage through the seal water collection tank, operators would isolate 1WG033 or 1WG035. These design features and proceduralized actions ensure that condenser vacuum is maintained. Thus, defeating the automatic isolation feature of 1WG074 would not significantly affect the frequency ofloss of condenser vacuum. The ANS Condition II classification ofthe transient is not affected because the frequency ofoccurrence is still moderate. Therefore, the proposed activity does not result in more than a minimal increase in the frequency of occurrence ofan accident previously evaluated in the UPSAR.

Evaluation No. 6G-17-004 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page2of2 Station/Unit(s): Byron I Unit 1 Activity/Document Number: EC # 619975 Revision Number: 0

Title:

TCCP to Install Pneumatic Jumper to Open 1WG074 The collection tank and level control system are not described in the UFSAR The condenser is described, but is not credited with mitigating the consequences of an accident or performing any other safety function. No other SSCs are affected by the proposed activity. Therefore, the proposed activity does not increase the likelihood ofoccurrence ofa malfunction ofan SSC important to safety previously evaluated in the UFSAR.

While loss of condenser vacuum is described as an initiating event, initiating events do not affect the consequence of an accident.

furthermore, the condenser is not credited with performing a safety function. Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UF$AR.

While loss ofcondenser vacuum is listed as an initiating event in UFSAR 15.2, the consequence ofa loss ofcondenser vacuum and its effect on SSCs important to safety is not changed. Therefore, the proposed activity does not result in an increase in the consequences of a malfunction of an $SC important to safety previously evaluated in the UFSAR.

Decrease in heat removal by the secondary is the only transient that could be initiated by a loss of condenser vacuum. Therefore, the proposed activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.

Malfunction of 1WG074 could potentially result in air inleakage to the condenser. Air inleakage and/or loss of condenser vacuum have already been evaluated in the UFSAR. Therefore, the proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.

Neither the FW pump seal water collection tank level control system nor the condenser are fission product barriers and do not interface with any fission product barriers either directly or indirectly. Therefore, the proposed activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

The Unit 1 Feedwater Pump Seal Water Collection Tank and associated level control equipment is not associated with any element of UfSAR described evaluation methodology used in establishing the design bases or used in the safety analyses. The proposed activity does not create a new or alternative evaluation methodology. Loss of condenser vacuum is an initiating event for, but does not affect the safety analyses for the decrease in heat removal by the secondary system transient. Therefore, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

The Feedwater Pump Seal Water Collection Tank and associated level control equipment are not addressed in the Technical Specifications or in the Operating License. Therefore, the Technical Specifications and Operating License do not require change.

Based on this evaluation, the proposed activity does not require prior NRC approval for implementation.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

Applicability Review 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 50.59 Evaluation No. 6G-17-004 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-17-005 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 1 of9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals Durini Normal Power Operation & Change WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description ofwhat the proposed activity involves.)

This activity will modify the Non-Safety Related Feedwater System (FW) Water Hammer Prevention System (WHPS).

Implementation ofthis activity encompasses the following.

(1) Provide the capability to disable the WFIP$ Steam Generator low level and low pressure feedwater isolation signals during normal operation at power levels that require the Main fW Isolation Valves 2FWOO9AIB/CID to be open.

The proposed change to disable the WFWS FW isolation involves installing a new switch in the circuit between two terminal block points for each of the feedwater Isolation Valves, 2fW009A/B/CID, and fW Preheater Bypass Valves, 2fW039A/BIC/D, and repositioning the control switch from Auto to Open for each ofthe FW Tempering Isolation Valves 2FW035A1B/C/D valves. The associated annunciator circuit and annunciator tile for 2FW035A/B/CID position is modified to reflect the new normal position for power operation (above approximately 25%). The WHPS FW isolation signals to the FW Isolation Bypass Isolation Valves 2fW043A/B/C/D valves are not disabled (these valves are normally in the closed position above 25% power level).

The following changes will be made to plant operating procedures to implement manual enabling/disabling of the WHPS FW isolation auto close signals:

. 2BGP 100-1, PLANT HEATUP, will be revised to verify that WHPS FW isolation control circuits are enabled prior to starting any FW pumps.

- 2BGP 100-3, POWER ASCENSION, will be revised to disable WFWS fW isolation after the 2FWOO9AJB/C/D valves are open.

. 2BGP 100-4, POWER DESCENSION, will be revised to enable WHPS FW isolation control circuits for the 2FW035A/B/C/D and 2FW039A/B/C/D valves prior to closing the 2FWOO9A/B/C/D valves, and then perform a follow-up action to enable WHPS FW isolation control circuits on the 2fWOO9AIB/C/D valves.

. 2BGP 100-4T4, REACTOR TRIP POST RESPONSE GUIDELINE, will be revised to verify that the 2FWOO9A/B/C/D, 2fW043A/B/C/D, and 2FW039A/B/C/D valves are closed prior to starting the Startup Feedwater Pump. Action will also be taken to enable WHPS fW isolation control circuits later in the post trip response.

UFSAR Section 10.4.7.3 will be revised as a result ofthis activity to describe the capability to disable the WHPS fW isolation control circuits when the Engineered Safety Features Actuation System (ESFAS) feedwater Isolation feature is active.

(2) Change the logic ofthe WHPS Steam Generator (SG) low level and low pressure control relays such that, when WFIPS is enabled, the WHPS actuation will occur when the relays are energized (versus currently de-energized).

As presently configured, the Non-Safety Related Feedwater (FW) Water Hammer Prevention System (WUPS) causes closure ofthe feedwater Isolation Valves (fWIVs) when control relays de-energize. The proposed activity will change this logic such that closure ofthe FWIVs will occur when the control relays are energized.

Implementation ofthis activity requires swapping wires from nomally open (NO) contacts to the normally closed (NC) contacts on existing WHPS relays that actuate on Steam Generator (SG) Level Low and SG Pressure Low conditions. Also, the existing Time Delay Relays related to WHPS SG Level Low Time Delay that currently start on a de-energized condition, will be replaced with new Time Delay Relays that start on an energized condition.

Evaluation No. 6G-J 7-005 50.59 REVIEW COVERSHEET FORM L$-AA-104-1001 Revision 4 Page 2 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Chance WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Energized To Actuate On Energized Additionally, this activity will functionally remove a second/redundant relay in the WHPS steam generator low pressure and low level permissive circuit for the FWWs (2FWOO9A/B/CID) that was added in 2005 under EC 348112 to eliminate a single point vulnerability the WHPS circuits caused by de-energization ofthe single relay in the original design. Upon installation ofthis EC 617266, WHPS actuation via this second/redundant relay is no longer needed since the previous single point vulnerability (WHPS actuation on de-energized condition) will no longer exist.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

This activity is being initiated following an unexpected loss ofpower to Non-Safety Related MCC 234V4, which subsequently caused loss of power to the WHPS control relays in cabinet 2PA281 related to 2B/2C SG Level Low and $G Pressure Low conditions. When the relays changed state from energized to de-energized due to the loss of power, the WHPS generated a signal to automatically close the 2B/2C fWIVs resulting in decreasing 2B/2C SG levels. In response to the inadvertent feedwater isolation and decreasing SG levels, operators manually tripped the reactor.

Implementation ofthe proposed change to provide the capability to disable the WHPS Steam Generator low level and low pressure FW isolation control circuits during power operation above start-up levels, will address an adverse vulnerability which can cause spurious/unnecessary WHPS actuation, feedwater isolation, loss ofnormal feedwater flow, and reactor trip solely due to the loss of a single non-safety related power supply for the relays. The changes to the plant operating procedures are made to control when the WHP$ FW isolation circuits are disabled/enabled.

Implementation ofthe proposed change to the control relay logic ofthe WHPS Steam Generator (SG) low level and low pressure control relays such that WHPS actuation will occur when the relays are energized (versus currently de-energized),

will prevent similar reactor trips in the future when WFJPS is enabled.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

(1) The safety function ofthe Feedwater Isolation Valves is to close, isolating main feedwater flow upon a safety injection signal, feedwater isolation signal and high-2 $G level. As described in UFSAR section 10.4.7.3, Several water hammer prevention features have been designed into the feedwater system. These features are provided to minimize the possibility of various water hammer phenomena in the steam generator preheater, steam generator main feedwater inlet piping, and the steam generator upper nozzle feedwater piping. Steam generator (two-out-of-three logic) low level trips are provided to close all feedwater isolation valves (2fW009A/B/C/D), feedwater isolation bypass valves (2FW043A1B/C/D) and feedwater preheater bypass valves (2FW039A/B/C/D). Steam generator (two-out-of-three logic) low pressure trips are provided to close all feedwater isolation valves (2FWOO9A/B/C/D), feedwater isolation bypass valves (2fW043A/B/C/D), feedwater preheater bypass valves (2FW039A/B/C/D), and the feedwater bypass tempering valves (2FW035A/B/C/D).

The UFSAR described water hammer prevention features were included in the NRCs basis for approval of(1) the designs ofthe Steam Generator and the FW/AFW systems, and the measures to address Unresolved Safety Issue (USI)

A-i Steam Generator Water Hammer; and (2) the removal of pipe restraints resulting from the elimination of arbitrary intermediate breaks in the fW system piping system design.

The purpose ofthe WHPS FW isolation is to prevent water hammer by precluding the admission of cold FW into a steam-voided preheater section ofthe steam generator (when SG narrow range level is less than 5% or when SG pressure is less than 600 psig). When the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals function to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. Note that the 2FW039 valves close on reactor trip and no coincident Tave is required. At low power levels or during startup the WHPS FW isolation features will be enabled. The design bases function to preclude the admission of cold FW into a steam-voided preheater section ofthe steam generator is maintained.

Evaluation No. 6G-1 7-005 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 3 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation SinaIs During Normal Power Operation & Chance WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized The other UFSAR described WHPS features will not be altered by this activity. Operating procedures associated with only providing feedwater to the upper nozzle ofthe steam generator during startup and low load conditions are unchanged. Temperature monitoring and alarms to detect possible back leakage of steam from the steam generators into the feedwater piping are unchanged. forward flushing and temperature permissive interlocks associated with opening the fWIV are also unchanged. This activity does not alter split feedwater flow operation.

Implementation ofthis activity does not impact the safety related circuits described in the UFSAR and Technical Specifications that generate an independent feedwater Isolation Signal, including Safety Injection (SI), SG Level HI-2 Trip (P-14), and Reactor Trip (P-4) coincident with RCS Average Temperature Low at 564°f. Additionally, implementation of this activity does not impact the Reactor Trip signal generated on low-low steam generator water level Implementation ofthis activity will require new operator actions to appropriately enable/disable WHPS FW isolation in the plant startup, power ascension, power descension, and plant trip response procedures. Operators will reposition the control board hand switch for the 2fW035 A/B/C/D valves and reposition the switches for the 2FW009A/B/C/D and 2FW039A1B/C/D valve circuits.

. for startup WHPS FW isolation will be verified to be enabled prior to starting any fW pumps. Once the 2FW009 valves are open, action will be taken to disable WHPS FW isolation for the 2FW009, 2FW035, and 2FW039 valves.

. For a normal shutdown or low power operation action will be taken to re-enable WHPS FW isolation for the 2fW035 and 2fW039 valves prior to closing the FW009 valves. WHPS FW isolation for the 2FW009 valves will be enabled prior to defeating the ESFAS fW isolation signals.

. On a plant trip action will be taken to verify that the 2fW009, 2FW039, and 2FW043 valves are closed prior to starting the Startup FW pump. The 2fW035 valve flow path is used to direct FW to the upper nozzle. The UFSAR described main FW flow and temperature permissive interlocks will prevent inadvertent introduction of cold FW to the lower nozzle until action is taken to enable WHPS FW isolation.

There is no impact on any emergency operating procedures. The proposed operator actions are not time critical and do not impose a significant new burden on the operating crew.

(2) This activity does not impact the independent Safety Related circuits described in the UFSAR and Technical Specifications that also generate a feedwater Isolation Signal, including Safety Injection (SI), SG Level HI-2 Trip (P 14), and Reactor Trip (P-4) coincident with Low RCS Average Temperature at 564°f.

The present logic configuration can cause feedwater isolation even though the SG Level and SG Pressure conditions necessary for WFWS actuation do not exist. Modifying the WHPS logic will prevent spurious actuation when WHPS is enabled and prevent a unit trip merely as a result of loss-of-power to control relays. The existing configuration causes the WHPS to actuate on loss ofNon-Safety Related power to the control relays. This power fail-safe design is not required for the WHPS in the same manner as that required for the independent Safety Related protection circuits stated above. There are no UfSAR or design requirements for the WHPS to actuate due to a loss ofpower condition in this manner. The existing design has demonstrated a vulnerability that results in a significant plant transient and reactor trip.

The modified design to change the logic ofthe WHPS Steam Generator (SG) low level and low pressure control relays such that WHPS actuation will occur when the relays are energized (versus currently de-energized), will eliminate this vulnerability.

Evaluation No. 6G-1 7-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 4 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Change WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On fnerized Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

50.59 Screening WHPS Disable Capability; The UPSAR described water hammer prevention features were included in the NRCs basis for approval of(1) the designs ofthe Steam Generator and the FW/AFW systems, and the measures to address Unresolved Safety Issue (USI) A-i Steam Generator Water Hammer; and (2) the removal of pipe restraints resulting from the elimination of arbitrary intermediate breaks in the FW system piping system design. As such, implementation of the capability to disable the WHPS Steam Generator low level and low pressure feedwater isolation signals j considered adverse and will be reviewed in the 50.59 evaluation.

WHPS Control Logic Change; The present logic configuration can cause feedwater isolation even though the SG Level and SG Pressure conditions necessary for WHPS actuation do not exist. Modifying the WHPS logic will prevent spurious actuation and prevent a unit trip merely as a result of loss-of-power to control relays. The modified logic will continue to provide the same Non-Safety Related WHPS function as described in the UFSAR. This function ofthe WHPS is Non-Safety Related and is independent ofthe Safety Related Engineering Safety Features (ESF) functions that also isolate feedwater. The WHPS system does not have a UfSAR described design function to isolate feedwater under any condition other than the low SG Level and low SG pressure conditions as described above. Additionally, the UFSAR does not discuss whether the WHPS is de-energize or energize to actuate. As presently configured, during a loss of power while at normal operation, the WHPS system will lose power and the relay logic will isolate feedwater from the SGs. This modification will eliminate this response on loss of power. However, a reactor trip on low SG level will occur prior to reaching the WHPS conditions of<5% SG Level and a Safety Injection actuation will occur on low SG pressure prior to reaching the WHPS conditions of <600 psig SG pressure. A reactor trip concurrent with low RCS Average Temperature or a Safety Injection actuation will close all FWIVs as required, independent ofthe WHPS. The independent feedwater isolation signals that occur following loss of power are safety-related. This activity does not impact the UfSAR described design function. The only circuit being altered is the FW WHPS circuit and the new design will function in the UFSAR described manner. The solenoids associated with the FWIVs are unchanged and will remain Energize to Close, such that system reliability is not impacted. Modification testing will ensure that the FWIV closure circuit design function remains as described in the UFSAR. The devices being added (new time delay relays) meet the same design, performance and installation standards as the existing devices in the circuit. All 1E to non- 1E circuit interfaces remain intact as well as train/non-train separation requirements as described in the UFSAR. In summary, this activity has been reviewed and determined to not have any adverse impacts to an UFSAR described design function.

WHPS Disable Capability; Implementation ofthis activity will require new operator actions to appropriately enable/disable WHPS FW isolation in the plant startup, power ascension, power descension, and plant trip response procedures. This change j considered adverse and will be reviewed in the 50.59 evaluation.

WHPS Control Logic Change; There are no procedure impacts with this change as the solenoids associated with the FWIVs are unchanged and will remain Energize to Close. The UfSAR does not describe the logic mechanism for how the WHPS signals are sent to valves. As such, this change is considered adverse. In summary, this activity has been reviewed and determined to not involve a change to a procedure that adversely affects how UPSAR described SSC design functions is performed or controlled.

There are no evaluation methodologies or alternative evaluation methodologies involved in this activity. This activity modifies the Unit 2 WHPS low SG level and low SG pressure FW isolation circuits to provide a means to disable the auto close signals when the unit is operating above approximately 25% power. This activity does not involve a method of evaluation as defined in LS-AA-104-1000. This activity will also modify the WHPS control logic so that WHPS actuation will occur when the relays for SG Level Low or SG Pressure Low are energized (versus currently de-energized), thus alleviating the potential for an inadvertent FWIV closure signal. Feedwater isolation from the WHPS is not related to any

Evaluation No. 6G-17-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 5 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Change WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized UPSAR described evaluation methodology. The logic mechanism for how the Non-Safety Related Feedwater WHPS signals are sent to valves requiring closure is not specified or referenced in the UFSAR and is not related to any UFSAR described evaluation methodology or used in establishing the design bases or safety analyses. Implementation ofthis activity does not impact the independent Safety Related Engineering Safety Features (ESF) FW isolation signals (Steam line pressure SI or the Reactor trip + Lo Tave) that also isolates feedwater. Therefore, this activity does not involve an adverse change to an element of a UFSAR described evaluation methodology, or use of an altemative evaluation methodology, that is used in establishing the design bases or used in the safety analyses.

The FW WHPS remains equivalent in function when feedwater isolation is appropriate on Low SG Level and Low SG Pressure. The revised design does not impact or interact with any other plant design functions. This activity requires post modification installation testing. Per LS-AA-104-1000, Section 4.2.2, Post-modification testing ofapproved facility changes is indistinguishable, in terms oftheir risk impact on the plant, from maintenance activities that restore SSCs to their as-designed condition. As such, installation and testing of approved facility changes are maintenance activities that must be assessed and managed in accordance with 1OCfR5O.65(a)(4), and therefore, this is not a test or experiment. The installation and testing activities will be structured to preclude interaction with other plant equipment and functions. Therefore, this activity does not involve a test or experiment not described in the UFSAR, where an SSC is controlled in a manner outside the reference bounds of the design.

The Non-Safety Related WUPS is not described in the Technical Specifications or Operating License and does not support the bases for any Technical Specification. The Technical Specifications Section 3.3.2 Engineering Safety Features (ESF) functions ofthe Feedwater Isolation Valves to close is not being changed in any way nor impacted by this activity.

Therefore, this activity has no impact on the Technical Specifications or Operating License and no Technical Specification or Operating License changes are required.

50.59 Evaluation Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not introduce the possibility of a change in the frequency of an accident because WHPS is not an initiator of any accident and no new failure modes are introduced. Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power reduces the probability of a loss of feedwater event due to a spurious WHPS actuation.

In the plant operating conditions under which the WHPS automatic FW isolation feature may be procedurally disabled (power level above approximately 25% power), the safety related ESFAS signals function to close the feedwater isolation valves on a low steam line pressure safety injection (< 640 psig) or steam generator water level low-low reactor trip (; 36.3% NR) coincident with low Tave (< 564° F). Note that the 2fW039A/B/CID valves close on reactor trip and no coincident Tave is required. A review of the postulated transients that could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 617266). The review determined that the ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip + Lo Tave ) will function to prevent water hammer by precluding the admission of cold FW into a steam- voided preheater section of the steam generator during power operating conditions. Therefore disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above 25% power does not increase the probability ofwater hammer in the SG preheater and the main FW inlet piping. Thus the proposed change does not result in more than a minimal increase in the frequency of occurrence of an accident caused by a water hammer event in the FW line or the Steam Generator (FW line break or steam generator tube rupture).

The defeated WHPS FW isolation function is not used or credited to mitigate any design basis event, analysis, or Emergency Operating Procedure Operator Action. The credited automatic function is the ESFAS FW Isolation which is a completely separate circuit. Thus the ESFAS FW Isolation is not affected by this change.

The change adds new operator actions to disable and enable the WHPS low level and low pressure FW isolation auto close signals. failure to enable the WHPS low SG level and low SG pressure auto close signals for startup, low load,

Evaluation No. 6G-17-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 6 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Change WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized or increasing load conditions could potentially result in a malfunction of equipment important to safety, specifically prevention of water hammer in the steam generator preheater and/or steam generator main feedwater (FW) inlet piping. The likelihood of occurrence of a malfunction associated with the new operator actions is not more than minimal based on the following:

. The actions to disable/enable the WHPS low SG level and low SG pressure fW isolation auto close signals has been added to the appropriate plant procedures and operators will be trained on the revised procedures.

. The actions to disable/enable the WHPS low SG level and low SG pressure fW isolation auto close signals do not have a required time for completion. Above 25% power there is no specific time requirement for disabling the WHPS low SG level and low $G pressure FW isolation auto close signals. During normal power descension there are no time limitations to enable WHPS, as this is directed by procedure prior to manually blocking SG steamline low pressure Safety Injection during shutdown/cooldown. There is no impact on any emergency operating procedures. The proposed operator actions for WFIPS enable/disable during power ascension, power descension, and reactor trip are not time critical and do not impose a significant new burden on the operating crew.

. This change does not introduce credible errors during performance of the operator actions required for this change. Three actions are required to disable or enable WHPS following implementation ofthis activity.

0 Operating a single (new) switch for each 2FWOO9A/B/C/D valve circuit at the Auxiliary Safeguards Relay Cabinet, 2PA27 (Train A) or 2PA2$J (Train B). Manipulation of the switch is considered a routine action with no credible failure risk. Opening/closing the switch does not initiate a water hammer prevention action; it disables/enables the respective WHPS circuit to the associated valve solenoid which remains unchanged by this activity as Energize to Close, such that system reliability is not impacted.

0 Operating a single (new) switch for each 2FW039AfB/C/D valve circuit at the Auxiliary Safeguards Relay Cabinet, 2PA27 (Train A) or 2PA2$J (Train B). These switches will be wired across the original water hammer prevention actuation relay contacts in cabinets 2PA27 (Train A) or 2PA2$J (Train B), thereby allowing the capability to disable/enable the original contact actuation for the WHPS circuit to the associated valve solenoid which remains unchanged by this activity as Energize to Open, such that system reliability is not impacted. Manipulation of the switch is considered a routine action with no credible failure risk.

0 Operation ofthe existing Main Control Room control switch (2HS-FW261/262/263/264) for the 2fW035A/B/C/D valves. With implementation ofthis activity, the 2FW035A/B/C/D valves will be kept open by placing the valve control switches to the OPEN position, thus bypassing the water hammer prevention circuit. When this control switch is placed in the AUTO position the 2fWO35A/B/CID valves will automatically close when there is a Steam Generator Low Pressure (2 out of3) condition. This is a routine method ofconfiguration control by operators. The associated valve solenoid remains unchanged by this activity as Energize to Open, such that system reliability is not impacted.

Operation ofthe new switches for the 2FW009A/B/C/D and 2FW039A/B/C/D valve circuits, and operation ofthe existing Main Control Room control switch (2H5-FW261/262/263/264) for the 2fW035A/B/C/D valves will be controlled as part ofan approved procedure implemented as part ofthis activity. The switches to be installed and used, and hand switches to be manipulated per this activity will be clearly labelled per the Configuration Change to allow for clear and easy operator identification. Operation ofthe switches does not initiate a water hammer prevention actuation; therefore, there would be time to recover from unlikely actions such as blocking or enabling WHPs out ofthe expected sequence.

. The purpose ofthe WHPS FW isolation signals is to prevent water hammer by precluding the admission of cold FW into a steam-voided preheater section ofthe steam generator. A review ofthe postulated transients that could create conditions for a water hammer event when power is above 25% was performed (Refer to EC 617266). The review determined that the ESFAS FW isolation signals (Steam line pressure SI or the Reactor trip + Lo Tave ) will function to prevent water hammer by precluding the admission of cold FW into a steam-voided preheater section ofthe steam generator during power operating conditions. Thus the system function to preclude the admission of cold FW into a steam-voided preheater section of the steam

Evaluation No. 6G-1Z-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 7 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Chance WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized generator is maintained.

When the WHPS FW isolation closure signals are disabled, the UFSAR described automatic action to isolate feedwater to the preheater region ofthe Unit 2 SGs on low SG level or low SG pressure will be fulfilled by the safety related ESFAS FW isolation signals. Appropriate administrative controls will be put in place to enable the automatic WHPS FW isolation function during plant startup, power descension, and afier a plant trip when the ESFAS FW isolation is not available. Thus the proposed change does not permanently substitute manual action for automatic action for performing a UPSAR-described design function.

The WHPS low SG level and low SG pressure FW isolation auto close signals are classified as non-safety related. The WHPS FW isolation auto close signal for each individual SG is based on a two-out-of-three logic. The safety related ESFAS FW isolation signal on low steam line pressure is based on two-out-of-three logic in any one ofthe four steam lines. A reactor trip is actuated on two-out-of-four low-low SG water level signals occurring in any steam generator. A FW isolation signal is generated with the reactor trip coincident with a two-out-of-four Lo Tave signals. The safety related ESfAS fW isolation signals provide the same level of redundancy as the WHPS FW isolation signals. The safety related ESFAS components are considered to be more reliable than the non-safety related WHPS components.

The highest risk for water hammer events is during startup and low power operation when the ESFAS FW isolation signals may be bypassed and only the WHPS FW isolation signals would be available for water hammer protection.

This same level of redundancy will be in place with only the ESFAS fW isolation signals providing protection when the WHPS low SG level and low SG pressure auto close fW isolation signals are disabled.

The new switches used to disable water hammer protection on the 2FW009A/B/C/D and 2FW39A/B/C/D valves meet or exceed the quality requirements ofthe circuits they are installed in and are installed in accordance with approved quality controlled procedures to ensure the appropriate level of reliability. The switches for the 2FWOO9A/B/C/D valves are installed in series with the applicable contacts with lugs on the terminal block and are considered to be as reliable as the existing circuit. The switches on the 2FW039A/B/CID valves are installed in parallel with the applicable contacts with lugs on the terminal block and are considered to be as reliable as the existing circuit. Existing procedures for Power Ascension (2BGP 100-3), Power Descension (2BGP 100-4) and Low Power Turbine Trip (2BOA TG-8) provide the necessary guidance for operator alignment ofthe FW system, prior to potential actuation of the WHPS, to prevent potential water hammer events. Byron Emergency Response Procedures (e.g. 2BEP ES-O. 1, Reactor Trip Response) already provide the necessary guidance for restoration of water flow to the upper SG nozzle and isolation of the main FW nozzle and preheater section per the water hammer prevention guidance provided by Westinghouse Emergency Response Guidelines. Thus the proposed change does not change the likelihood of occurrence of a malfunction of the WHPS to preclude a water hammer event The Main Control Room control switches (2HS-FW26 1/262/263/264) for the 2FW035A1B/C/d valves are designed to allow the valves to be held open. Thus the proposed change does not change the likelihood ofoccurrence of a malfunction ofthese valves. The associated annunciator circuit and annunciator tile is modified to reflect the new normal position for power operation (above approximately 25%).

The proposed change does not impact the other UFSAR described WHPS features. When the WHPS FW isolation is disabled at power, the water hammer prevention isolation function to preclude injection of cold water into a steam-voided preheater section ofthe steam generator will continue to be fulfilled by the safety related ESFAS signals to close the feedwater isolation valves on low steam line pressure safety injection or steam generator water level low-low reactor trip coincident with Tave less than 564° F. At low power levels or during startup the WHPS FW isolation features will be enabled.

UFSAR Section 10.4.7.3 will be revised as a result ofthis activity to describe the capability to disable the WFIPS FW isolation control circuits when the Engineered Safety Features Actuation System (ESFAS) Feedwater Isolation feature is active.

Additionally, the proposed change does not revise the operating procedure steps associated with switching feedwater flow from the top feed nozzle to the lower main feed nozzle. As described in the response to FSAR Question 010.51 initial plant startup testing was performed to confirm that no damaging water hammer occurs when FW delivery is transferred from the top SG nozzle to the main feed nozzle. Since the transfer procedure steps are unchanged, there is no change in the frequency of a water hammer when FW delivery is transferred. Therefore the proposed change will not result in more than a minimal increase in the likelihood of a water hammer induced malfunction.

Evaluation No. 6G-1Z-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 8 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals Durini Normal Power Operation & Chance WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized Based on the above, the proposed activity will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an $$C important to safety previously evaluated in the UPSAR.

Disabling the WHPS low SG level and low SG pressure FW isolation auto close signals above approximately 25% power does not introduce the possibility of a change in the consequences of an accident because the fW isolation from the WI-PS is not credited with mitigating any accident previously evaluated in the UfSAR. The ESFAS fW isolation signals (Steam line pressure SI or the Reactor trip + Lo Tave ) will continue to close the feedwater isolation valves as designed for accident mitigation. Thus the proposed activity will not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.

Disabling the WI-PS low SG level and low SG pressure fW isolation auto close signals above approximately 25% power does not introduce the possibility of a change in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR because the FW isolation from the WI-PS is not credited to mitigate the consequences of malfunctions of an SSC important to safety. The proposed change does not impact the safety related ESFAS FW isolation circuitry. Thus there is no change to the UPSAR described failure modes and effects for the ESF actuation system. The proposed changes to the WI-PS fW isolation auto close signals does not introduce any new failure mode for the system Disabling the WI-PS low SG level and low SG pressure FW isolation auto close signals above approximately 25% power does not introduce the possibility of a new accident because the proposed change is not the initiator of any accident and no new failure modes are introduced. When the WI-PS FW isolation is disabled at power the automatic water hammer prevention function will be fulfilled by the safety related ESFAS signals for FW isolation. The isolation offW in events resulting in a reduction of FW temperature without a reactor trip (e.g., turbine trip below P-8 permissive) was never an automatic feature in the WI-PS. As per original design, operators will be notified of low fW temperature or low FW flow by Main Control Room annunciators and manual operator action will be taken to correct the condition or to isolate FW flow to the SG preheater section per existing BAR procedures (Ref. BAR 2-15-Al 1, BAR 2-15-Al2, et.al.). This manual operator action to address low FW temperature or flow is not affected by this activity. Administrative controls will be put in place to ensure that the WHPS fW isolation signals are enabled during startup and low power operation. The likelihood of operator error leading to an accident of a different type is not credible because the actions are reflected in plant procedures and operator training programs, the actions can be completed in the required time, the actions are simple and adequate time is available to recover from any credible error. The design bases function to preclude the admission of cold FW into a steam-voided preheater section of the steam generator is not adversely impacted, thus no new accident types are introduced.

Disabling the WI-PS low SG level and low SG pressure fW isolation auto close signals above approximately 25% power does not introduce the possibility of a malfunction with a different result because when the WHPS FW isolation is disabled at power the water hammer prevention function will be fulfilled by the safety related ESFAS signals. Administrative controls will ensure that the WI-Ps FW isolation signals are enabled during startup and low power operation This proposed modification ofthe Unit 2 WI-PS low SG level and low SG pressure FW isolation auto close signals does not result in a change that would cause any system parameter to change. WHPS is not credited for mitigation in any accident analysis, thus there is no impact on any UFSAR described fission product barrier limit. Therefore, the proposed activity does not result in a Design Basis Limit for a fission Product Barrier (DBLfPB) as described in the UFSAR being exceeded or altered This activity modifies the Unit 2 WI-PS low SG level and low $G pressure FW isolation circuits to provide the capability to disable the auto close signals when the units are operating above approximately 25% power. The proposed change does not involve a method ofevaluation as defined in LS-AA-104-l000. Therefore, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analysis.

Evaluation No. 6G-J 7-005 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 9 of 9 Station/Unit(s): Byron Unit 2 Activity/Document Number: EC 617266 I DRP 17-023 Revision Number: 000 I 0

Title:

Provide Capability to Disable Water Hammer Prevention System (WHPS) Feedwater (FW) Isolation Signals During Normal Power Operation & Change WHPS SG Low Pressure/Level Control Relay Operation From Actuate On De-Enerized To Actuate On Energized Therefore, based on this 50.59 Screening and Evaluation, this activity may be implemented under the governing station procedures without prior NRC approval.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

U Applicability Review 50.59 Screening 50.59 Screening No. 6E-17-042 Rev. 0 50.59 Evaluation 50.59 Evaluation No. 6G-17-005 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-17-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 1 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Change # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS EAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown NOTE: for 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CfR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description ofwhat the proposed activity involves.)

The proposed activity involves revision of the licensing and design basis for safety significant components that are not fully protected from tornado-generated missiles. The change involves the following:

1. Analyses that establish missile damage acceptance criteria for the exposed Ultimate Heat Sink (UHS) riser pipes, the exposed 1B/2B Auxiliary Feedwater (Af) diesel engine exhaust, the Main Steam (MS) PORV exhaust tail pipes, and the Miscellaneous Electric Equipment Room (MEER) exhaust openings in the wall between the Auxiliary Building and the Turbine Building (U-line wall). The missile damage analyses for UHS riser pipes and the MS PORV exhaust tail pipes are dispositioned in 50.59 Screening No. 6E-17-047. The missile damage analyses for the 1B/2B Auxiliary Feedwater (AF) diesel engine exhaust and the MEER exhaust openings are dispositioned in 50.59 Evaluation No. 6G-17-004.

2. The Byron tornado missile analysis using the TORMIS methodology. TORMIS uses a Monte Carlo simulation technique to assess, through a Probabilistic Risk Assessment (PRA) methodology, the probability of multiple missile hits causing unacceptable damage to unprotected safety-significant components at a plant. The results from the Byron TORMIS analysis are used to credit the unprotected equipment for post-tornado safe shutdown. A License Amendment Request was submitted and approved to utilize the TORMIS computer code methodology. Thus, this part ofthe activity is dispositioned in the 50.59 Applicability Review.

3 Revised UHS temperature analysis for safe shutdown from a tornado event. The Essential Service Water Cooling Tower (SXCT) fans and cells that are shown via the TORMIS analysis to survive a tornado strike will be credited for UHS cooling. The revised UHS temperature analysis for safe shutdown from a tornado is dispositioned in 50.59 Evaluation No. 6G-17-004.

4. New TRM 3.7.e, Tornado Design Basis SXCT fans Operating and TRM 3.7J Tornado Design Basis SXCT fans Shutdown are put in place to ensure that the assumed initial conditions in the post-tornado UHS cooldown analysis are met; i.e. the administrative controls will specify the number of SXCT fans required to be operable based on outside environmental conditions. The new TRMs are dispositioned in 50.59 Screening No. 6E-17-047.

5. Procedure changes to implement the new TRMs and the operator actions assumed to be performed in the supporting analyses. The specific operator action changes are listed in the Effect ofActivity section ofthis 50.59 Review Coversheet Form. Procedure changes to implement the new TRMs are dispositioned in 50.59 Screening No.

6E-17-047. The procedure changes for the operator actions are dispositioned in 50.59 Evaluation No. 6G-17-004.

6. UfSAR Changes associated with the TORMIS method and analysis. The TORMIS UFSAR changes are dispositioned in the 50.59 Applicability Review.

7. UFSAR Changes associated with the revised analysis of the UHS for safe shutdown in a tornado event. These UFSAR changes are dispositioned in 50.59 Screening No. 6E-17-047.

8. UFSAR Changes to correct the descriptions associated with the following:

Evaluation No. 6G-f 7-006 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 2 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Change # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS LAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Desiin Basis SXCT Fans Shutdown

. UFSAR Section 3 .5.3 The description the walls and roofs of structures protecting components from tornado generated missiles is revised to replace the minimum thickness values with a statement that the thicknesses are well above the calculated missile penetration thicknesses.

. UfSAR Section 2.4.1 1 .6 is revised to indicate that the onsite wells are located in reinforced concrete enclosures as opposed to the current description of located in enclosed heated buildings to provide freeze protection.

. UFSAR Sections 3.5.4.2 and 9.5.8.2 are revised to remove the statement: Analysis has established that the stacks can be damaged to the extent that the flow area is reduced to 50% of the original flow area without reducing the diesel power output.

These non-editorial UFSAR changes are dispositioned in 50.59 Screening No. 6E-17-047.

9. Editorial UFSAR Changes associated with the description oftornado winds, tornado missiles, tornado missile protection features, and the UHS. The editorial UFSAR changes are dispositioned in the 50.59 Applicability Review.

10. Procedure changes to remove the compensatory actions put in place associated with for equipment identified as not adequately protected from tornado missiles under Enforcement Guidance Memorandum 15-002. The subject equipment was included in the Byron TORMIS analysis. With the NRC approval of TORMIS and the change to the licensing basis, the EGM 15-002 compensatory actions are no longer required. These procedure changes are dispositioned in the 50.59 Applicability Review.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

In NRC Inspection Report, Byron Station, Units 1 and 2, Integrated Inspection Report 05000454/2009004; 05000455/2009004, dated November 5, 2009, Byron Station received a non-cited violation, NCV, for failure to protect the Emergency Diesel Generator (EDG) Diesel Oil Storage Tank (DOST) vent lines from tornado-generated missiles. During the extent of condition review to address this violation, additional safety related pipes vulnerable to tornado-generated missiles were identified (i.e., the Steam Generator Power Operated Relief Valve (PORV) tailpipes, the Main Steam Safety Valve (MSSV) tailpipes, and the 1B/2B Auxiliary Feedwater (AF) Diesel exhaust stacks).

Subsequently, in NRC Inspection Report, Byron Station, Units I and 2, NRC Component Design Bases Inspection; Inspection Report 05000454/2015008; 05000455/2015002 and Notice ofViolation, dated July 21, 2015, Byron Station received another non-cited violation, for failure to evaluate the adverse effects of changing the SXCT Tornado Analysis as described in the UFSAR, Section 3.5.4, Analysis ofMissiles Generated by a Tornado, Revision 14. This revision changed the UFSAR to assume that two SXCT fans survive a tornado strike. As a result ofthis violation, the UFSAR analysis ofrecord reverted back to the original licensing basis assumption that multiple tornado missile hits could result in the loss of all SXCT fans. An Operability Evaluation was completed and is currently in place to address the concern that Cold Shutdown cannot be achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no SXCT fans available afler a postulated tornado event. This Operability Evaluation will be closed out after implementation of this activity.

On May 25, 2016, Byron Station issued Event Notification 51958, Discovery of Non-Conforming Conditions During Tornado Hazards Analysis. This Notification Report documents non-conforming conditions in the plant design such that specific Technical Specification equipment on both units was considered to be inadequately protected from tornado missiles. These conditions were addressed in accordance with Enforcement

Evaluation No. 6G-1 7-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 3 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Chanie # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS LAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis $XCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown Guidance Memorandum 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance, dated June 10, 2015 and DSS-ISG-2016-O1, Clarification ofLicensee Actions in Receipt of Enforcement Discretion Per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance.

To resolve the above concerns and close out the Operability Evaluation, Byron Station requested NRC approval to utilize the TORMIS Computer Code methodology for assessing tornado-generated missile protection ofthe Byron Station SSCs. Unprotected targets needed for safe shutdown after a tornado, including the unprotected UHS components, are included in the TORMIS analysis. The results ofthe TORMIS analysis specific to the post-tornado survival of the SXCT fans are then used to evaluate SXCT capability for safe shutdown after a tornado event.

NRC approval ofthe TORMIS methodology for Byron was granted via NRC Safety Evaluation dated 8/10/2017.

This activity implements the change in the licensing and design basis for safety significant components that are not fully protected from tornado-generated missiles. This includes revision ofthe UHS licensing basis, implementation ofnew TRMs and procedure changes that support the revised UHS analysis. The UFSAR change package revises the UFSAR to describe application ofthe TORMIS method, the revised UHS tornado analysis, and makes other UFSAR changes to clarify the description of the design basis for tornado winds, tornado missiles, tornado missile protection features, and the UHS.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

Design bases The activity revises the design bases associated with the effects oftornado and high wind generated missiles. The design is revised to utilize the TORMIS Computer Code as a methodology for assessing tornado missile protection of selected plant components. The TORMIS analysis utilizes a probabilistic approach to determine the acceptability of the current level of missile protection (or lack of missile protection) for the modeled plant components on each unit.

The TORMIS analysis is a supplement to the current licensing/design bases for tornado generated missiles. The original deterministic tornado generated missile design basis still applies to the barriers that provide missile protection to the systems, structures, and components not evaluated using TORMIS.

The original licensing basis assumed that multiple tornado missile hits could result in the loss of all SXCT fans. The application of the TORMIS analysis results revises the design basis for the UHS for post-tornado operation from natural draft cooling to mechanical draft cooling (i.e. from passive to active cooling tower fans).

Safety analyses described in the UFSAR The UFSAR described safety analysis for the UHS after the loss of SXCT fans due to tornado-generated missiles is revised to credit unprotected SXCT fans that survive a tornado strike based on the results ofthe TORMIS analysis.

The revised UHS analysis considers SXCT fans that may be out of service and a worst case single failure that results in the failure oftwo SXCT fans. Based on the results ofthe TORMIS analysis:

. 3 or more SXCT fans survive when one SXCT fan is out of service and two failed SXCT fans are assumed.

. 2 or more SXCT fans survive when two SXCT fans are out of service and two failed SXCT fans are assumed.

Evaluation No. 6G-17-006 50.59 REVIEW COVERSHEET FORM L$-AA-104-1001 Revision 4 Page 4 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Chanie # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS LAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown The number of SXCT cells that may be initially out of service is dependent on the outside air wet bulb temperature and number of operating units. The revised UHS analysis for a tornado event determined that with the available fans and risers, Cold Shutdown conditions can be reached for both units within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> while maintaining the SX supply water temperature less than or equal to the design maximum temperature of 1OO F.

Plant operations The activity has the following impact on plant operations:

. New operating limits are put in place associated with the number of required Operable SXCT fans based on the outside air wet bulb temperature. New required actions and completion times are established to ensure that the initial conditions assumed in the revised analysis are met.

. The existing time critical operator action to manually initiate the deep well pump(s) within 90 minutes of a LOCA is revised to include a tornado event where SX makeup from the Rock River is not available.

. The existing time critical operator action to isolate SX blowdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of a LOCA is revised to include a tornado event where SX makeup from the Rock River is not available.

. Operator action is required to isolate the AF telltale vent lines is required within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of a tornado event where $X makeup from the Rock River is not available. This action is currently directed in 1/2BEP ES-O.2, NATURAL CIRCULATION COOLDOWN, step 2 1 .b, if less than or equal to 4 SXCT fans are running with associated risers open.

. Operator action is required to isolate the SXCT riser leak-off drains within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of a tornado event that damages SXCT fans. This makes permanent the action put in place under Operability Evaluation 13-007.

This action is currently directed in 1/2BEP ES-0.2, NATURAL CIRCULATION COOLDOWN, step 21.b, ifless than or equal to 4 $XCT fans are running with associated risers open and in procedure BOP SX-T2, Step D.4.

. For post-tornado cooldown with damage to the SXCT, AF cooling concurrent with RH cooling will be used until the RCS is cooled below 280° F.

. For a post-tornado event shutdown, operators will use the existing procedural guidance on system temperature limits to control the rate of cooldown. With the postulated minimum number of SXCT fans (fans initially out of service, fans lost to a postulated single failure, and fans damaged by tornado missiles), the cooldown time may approach 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

. For a tornado event with missile damage to the MEER exhaust, manual operator action is required within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore ventilation cooling to the affected room(s). This action is currently directed in OBOA ENV 1, Attachment C, Step 2.

Plant operating procedures and operator training programs will be revised to reflect the new or modified operator actions. The following procedures are revised/created:

. OBOL 7.e and OBOL 7.f, new procedures associated with the new TRMs.

. OBOA ENV-1, ADVERSE WEATHER CONDITIONS UNIT 0

. BOP SX-T2, SX TOWER OPERATION GUIDELINES

. OBOSR 0.1-0, UNIT COMMON ALL MODES/ALL TIMES SHIFTLY AND DAILY OPERATING SURVEILLANCE

. OP-BY-102-106, OPERATOR RESPONSE TIME PROGRAM AT BYRON STATION

Evaluation No. 6G-JZ-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 5 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Chanie # 17OO4g DRP # 16030 RCVISjOfl Number: 0/0/0

Title:

PROCESS TORMIS EAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis $XCT Fans Operating and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown Summary of Conclusion for the Activitys 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

In accordance with 10 CFR 50.90 Byron Station requested a license amendment to utilize the TORMIS Computer Code method for assessing tornado-generated missile protection of unprotected plant systems, structures, and components. The LAR stated that the results from the Byron Station TORMIS analysis will be used to credit unprotected equipment for post-tornado safe shutdown. The LAR specifically indicated that the Essential Service Water Cooling Tower (SXCT) fans and cells that survive a tornado strike will be credited for Ultimate Heat Sink (UHS) cooling as opposed to the original licensing basis that assumed all the unprotected SXCT fans are damaged by tornado-generated missiles. The Safety Evaluation dated 8/10/2017 (ML17188A155), approved the requested licensing basis change. Thus, this portion ofthe proposed activity is controlled under 10 CFR 50.90 and a 50.59 review is not required as noted in the 50.59 Applicability Review.

A number of the proposed UfSAR changes involve reformatting, simplification, and editorial changes. These changes screened out as a Non-Regulatory Change in accordance with the provisions ofNEI 98-03, Revision 1 as noted in the applicability review.

Tornado missile damage to the UHS riser pipes that causes less than 50% crimping does not adversely affect performance ofthe UHS or Essential Service Water (SX) systems. The Essential Service Water Cooling Tower (SXCT) riser pipes are part ofthe SX system pressure boundary and function to provide a flow path from the underground SX return piping to the SXCT fill distribution piping. Tornado missile damage that causes pipe crimping does not result in loss of the SX system pressure boundary, thus this design function is not adversely affected. Piping flow analysis shows that 50% riser crimping has a negligible impact on system flow rates to components cooled by the SX system and negligible impact to the flow rates to the SXCT fill. The UHS temperature analysis included a sensitivity case to show that the revised flow rates with riser pipe crimping do not adversely impact the cooling function.

Tornado missile damage to the Main Steam (MS) Power Operated Relief Valve (PORV) exhaust tail pipes that causes less than 90% crimping does not adversely affect the relief valve function to relieve steam and maintain the steam generator pressure below the design value. Flow analysis shows that the allowed exhaust pipe crimping has no effect on the valve relieving capacity.

The damage acceptance criteria for tornado missile damage to the 1B/2B AF Diesel engine exhaust does not adversely affect the function ofthe 1B/2B AF pump, 1B/2B Af cube cooler fan, SX booster pump, the lube oil and the engine fuel oil pumps.

The proposed non-editorial UFSAR changes do not affect the function of any system, structure or component. The UPSAR changes to describe the TORMIS method and analysis reflect the approved licensing basis. The UFSAR changes associated with the revised UHS temperature analysis for a tornado event reflects the revised design basis analysis.

The UFSAR Section 3.5.3 description ofthe minimum wall thickness for the concrete walls and roofs of structures protecting the safety related systems and components from tornado-generated missiles is revised to remove the specific concrete thickness values. Design analyses show that the various concrete barrier thicknesses are well above the thickness required to prevent spalling and scabbing and the revised description conforms to the requirements of

Evaluation No. 6G-17-006 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 6 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity[Document Number: EC 400500, TRM Change # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS EAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown Section 3.5.3 ofRegulatory Guide 1.70, Revision 2. Thus, the proposed UFSAR change to remove the barrier thicknesses has no impact on the UFSAR described missile barrier function.

The proposed UFSAR change to indicate that the onsite wells are located in reinforced concrete enclosures as opposed to located in enclosed heated buildings does not adversely affect the function ofthe deep well pumps. The UFSAR described design functions are not affected by this UFSAR change because the deep well pump are located below grade in a reinforced concrete enclosure. No heating of the enclosure is required because the well head and discharge piping are located below grade which prevents freezing. The deep well pump enclosures were included as targets in the TORMIS analysis. Based on the results of the TOR1VIIS analysis, the deep well pump enclosures provide an adequate level of protection from tornado missiles and may be credited as an alternate UHS makeup water source after a tornado event.

The proposed UFSAR change to remove the statement that the Emergency Diesel Generator (EDG) exhaust stacks can be damaged to the extent that the flow area is reduced to 50% ofthe original flow area without reducing the diesel power output does not affect the function ofthe EDGs. As described in UfSAR Section 3.5.4.2, a rupture disc pressure relief device is installed on each diesel exhaust line. This relief device is located downstream of the silencer and inside the missile protection structure on the roof ofthe auxiliary building. Upon blockage of the stack, the rupture disc will open prior to backpressure increasing to the point that required diesel power is not available. The emergency diesel generators will therefore remain functional following any postulated tornado missile impact. Since the EDG exhaust system is designed to mitigate blockage of the stack, the description ofthe impact of a 50% flow area reduction is extraneous information that may be removed from the UfSAR.

The analysis to determine the acceptable amount oftomado missile damage for the UHS riser pipes, 1B/2B AF diesel engine exhaust, and MS PORV exhaust tail pipes does not involve a change to how UFSAR described design functions are performed and controlled. The analyses determine the allowed amount of tornado missile damage to these components. The allowed crimping (reduction in flow area) does not change how the UHS, 1B/2B AF diesel engine, or MS PORVs functions are performed or controlled. Thus, the analyses do not involve a change that adversely affects how a UFSAR described design function are performed or controlled.

The UFSAR changes associated with the revised URS temperature analysis for a tornado event reflects the revised design basis analysis. Thus, the proposed UFSAR changes do not adversely affect how UFSAR described SSC design functions are performed or controlled.

The change involves flow analysis for the UHS, 1B/2B AF diesel engine exhaust, and MS PORV exhaust. The UFSAR does not describe the method of evaluation for determining flow in the UHS, 1B/23 AF diesel engine exhaust, or MS PORV discharge lines. Thus, the change does not involve an adverse change to an element of an UFSAR described evaluation methodology.

The change involves temperature analysis ofthe MEER associated with the potential loss of room ventilation.

UFSAR Section 3 1 1.4.2 does not describe the method used to determine MEER room temperatures for a loss of ventilation event. Thus, the change does not involve an adverse change to an element of an UFSAR described evaluation methodology.

The change involves the UFSAR description ofthe minimum thickness of concrete barriers used for tornado missile protection of safety-related systems and components. As described in UFSAR Section 3 .5.3, the depth of penetration into a concrete barrier is calculated using the modified Petry equation and that the concrete barriers are designed such that the missile penetrates no more than two-thirds of the thickness of the barrier thus preventing spalling or

Evaluation No. 6G-17-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 7 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Change # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS EAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown scabbing. The proposed UFSAR change does not revise, change, or utilize an alternate method for determining the adequacy ofthe concrete barriers to prevent spalling and scabbing. Thus, the proposed change does not involve an adverse change to an element of a UfSAR described evaluation methodology.

The proposed activity changes the UfSAR described operation ofthe UHS after a tornado event. The change revises the evaluation to credit forced draft cooling as opposed to operation in strictly natural draft cooling. Use of forced draft cooling ofthe UHS is currently evaluated for normal and post-LOCA operation. UHS temperature analysis for safe shutdown from a tornado event shows that maximum supply SX water temperature remains within the UFSARJSER described maximum temperature and supports safe shutdown of both units within the UFSAR described time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reach cold shutdown. Thus, the change does not result in operation ofthe UHS or the sx system in a manner inconsistent with the descriptions in the UFSAR.

The new tornado design basis operating limits associated with the number of required Operable SXCT fans based on the outside air wet bulb temperature are consistent with current operating limits currently in Technical Specification 3.7.9. Thus, the new TRM limitations do not result in operation or control ofthe system in a different manner.

Operator actions to maintain UHS inventory and maximize UHS heat removal during a tornado event are consistent with current operating procedures for the UHS during a LOCA event. Existing procedural guidance is used to control the cool down rate and heat input to the UHS. Thus, the proposed changes do not operate the UHS, SX system, or equipment used for safe shutdown in a manner that is different that already described in the UfSAR.

The proposed change evaluates operation ofthe equipment located in the Miscellaneous Electrical Equipment Rooms after a tornado event for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without room ventilation. UFSAR Section 3 .1 1.4.2 and Table 3 1 1-2 currently describe MEER room heat-up for a postulated HELB event. The UfSAR describes that room equipment is capable of operating with elevated room temperatures that may occur after a loss of room ventilation. Design analysis has been performed to show that the resultant peak room temperatures will not exceed the EQ temperature limits of the equipment. TRM 3.7.d, Area Temperature Monitoring, allows the room temperature in the MEER and Battery Rooms to exceed the normal maximum design temperature for a limited period of time. Thus, the proposed change does not result in operation of a SSC in a manner that is inconsistent with the UFSAR description or outside the reference bounds of the design.

The proposed UfSAR changes, revised design analysis, new TRM operating limitations, and associated procedure changes do not require a change to Technical Specifications, because the changes do not alter or modify any existing Technical Specification.

The proposed changes associated with the analysis oftornado missile damage and the revised UHS temperature analysis for safe shutdown from a tornado event do not introduce the possibility of a change in the frequency of an accident because the revised analyses do not involve the initiator of any accidents. The proposed changes in operator actions mitigate the consequences of postulated tornado missile damage. The proposed operator actions are not the initiator of any accidents and are simple such that operator errors are judged to not be credible. The proposed changes do not impact the normal or accident function ofthe 1B/2B Af pump diesel engine, equipment in the MEER or Battery Rooms, or the UHS. The proposed activity is associated with the tornado missile impact on the 1B/2B Af diesel exhaust pipe and cover, postulated missile damage to the MEER and Battery Room exhaust ducts, and the UHS temperature analysis for safe shutdown from a tornado event. UFSAR changes will be made under DRP No.16-030 to describe the revised analyses. Thus, the proposed activity does not increase the frequency ofan accident previously evaluated in the UFSAR.

Evaluation No. 6G-17-006 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 8 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Change # 17OO4g DRP # 16030 Revision Number: 0/0/0

Title:

PROCESS TORMIS LAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operatin, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown Design analysis DO-EDS-1 determined the amount of crimping that would be acceptable for a tornado missile hit on the 1B/2B AF diesel engine exhaust pipes and exhaust cover plates. The allowed amount ofcrimping is used in design analysis ARA-0021 16, Tornado Missile TORMIS Analysis for Byron Generating Station. The allowed amount of crimping may result in higher engine exhaust back pressure. The higher engine exhaust back pressure with crimping was evaluated and could result in a slight reduction in the engine power output (O.65% reduction). There is approximately 12.7% margin between the full load on the 1B/2B AF diesel engine (Af pump full load HP, AF room cubicle cooler fan HP, AF lube oil and fuel pumps, and the SX booster pump HP) and the continuous break horsepower rating for the engine. Additionally, for safe shutdown from a tornado, the 1B/2B AF pump load on the engine will be reduced after a short time as SG level recovers and 1B/2B Af flow is throttled to match the heat removal rate via steaming. The potential 0.65 % reduction in 1B/2B AF diesel engine output due to missile damage on the exhaust pipe or cover plate has a negligible effect on the 1B/2B AF engine and the 1B/2B AF pump. Thus, the allowed amount oftornado missile hit crimping on the 1B/2B AF exhaust pipe or cover plate does not result in a more than minimal increase the likelihood of a malfunction of the IB/2B AF diesel.

for the MEER HVAC openings, the TORMIS model used pipe penetration as failure as opposed to a missile hit on the opening. Using this approach, TORMIS determined that the equipment within the MEER rooms could be credited for safe shutdown from a tornado. The dampers in the wall penetration could be potentially damaged by tornado missiles such that exhaust path is blocked and the affected MEER and Battery Room could loss ventilation cooling. Design Analysis BYR16-048 evaluates room temperatures during this loss ofventilation event. The analysis assumes that operator action will be taken within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to mitigate the loss of ventilation. Design Analysis EQC BB-008 previously evaluated the safety related equipment in the MEERs following a HELB event and determined that the components can withstand a 2-hour exposure of 165° F. BYRI6-048 determined that the MEER Division I maximum temperature at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would be 140° F and the MEER Division 2 maximum temperature at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would be 159° F. Since the calculated maximum MEER room temperatures remains below the evaluated temperature ofthe equipment in the room, the potential loss of ventilation does not adversely impact the function of the equipment important to safety located in the MEER rooms.

The post-tornado temperature excursion in the Battery Rooms was also evaluated. The calculated maximum peak Battery Room temperature was 1 19°F. The post-tornado increase in room temperature is within the limits of TRM.7.d, which allows the Battery Room temperatures up to 138° F for less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The loss of room ventilation for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would not result in an unacceptable build-up of hydrogen with the battery rooms as indicated in design analysis NED-H-MSD-017, Revision 4. The analysis concludes upon a loss ofventilation it takes 17.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reach 2% hydrogen concentration in the most limiting battery charge mode. Thus, the potential loss of ventilation does not adversely impact the function ofthe equipment important to safety located in the Battery Rooms.

Operator action is credited within two hours to mitigate the MEER and Battery Room heat-up if tornado missile damage results in loss room ventilation. The new action (including the required completion time) is added to plant operating procedures and operator training programs. The required manual action involves simple activities to open doors (to provide an exhaust path) and restart supply fans which can be completed within the required time.

Environmental conditions in the plant once the tornado/storm has passed would not challenge performance of this operator action. No credible errors are associated with performing this manual action. Thus, the new operator action would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety.

The revised UHS temperature analysis for safe shutdown from a tornado event credits new and existing operator actions to minimize the SX supply water temperature. The new or modified operator actions (including the required completion time) supporting the UHS temperature analysis are added to plant operating procedures and operator

Evaluation No. 6G-1 7-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 9 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500g TRM Chanie # 17OO4g DRP # 16030 Revision Number: 0/0/0

Title:

PROCESS TORMIS LAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown training programs. Based on validation by Operations, the required actions can be completed in the time required and the resources available. Environmental conditions in the plant and outdoors once the tornado/storm has passed would not challenge performance ofthe new operator actions. The proposed actions to start the deep well pump(s), isolate sx blowdown, isolate the AF telltale vent lines, and isolate the SXCT riser leak-off drains are simple such that operator errors and recovery actions are judged to not be credible. Operators already control the cool down rate during normal plant shutdowns to maintain RH, CC, and SX temperatures within operating limits. Thus, the new and modified operator actions for the UHS would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety.

The proposed changes associated with the analysis of tornado missile damage and the revised UHS temperature analysis for safe shutdown from a tornado event do not introduce the possibility of a change in consequence of an accident because the change involves the evaluation of safe shutdown after a tornado event. Based on the results of the approved TORMIS analysis, SXCT fans and cells that survive a tornado strike are credited for URS cooling. The post-tornado UHS temperature analysis considers a limiting failure and shows that with the available UHS and risers cold shutdown conditions can be reached for both units well before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while maintaining the SX supply water temperature less than or equal to the maximum design temperature of 1OO F.

The proposed changes in operator actions mitigate the consequences of postulated tornado missile damage. The proposed operator actions are not the initiator of any accidents and are simple such that operator errors are judged to not be credible.

Design analysis ofthe equipment important to safety located within the MEER and Battery Rooms and the AF Diesel Engine will continue to support safe shutdown with the missile damage acceptance criteria utilized in the TORMIS analysis. Thus, the consequences of a tornado event are unchanged. The proposed changes do not affect the operation of any S$C for other UFSAR accidents. Thus, there is no increase in the consequences of an accident previously evaluated in the UFSAR.

The revised UHS temperature analysis for safe shutdown from a tornado event credits SXCT fans and cooling tower cells that the TORMIS analysis indicates will survive a tornado event. The existing licensing basis assumed that all SXCT fans were damaged by tornado missiles and that the SXCT operated strictly in a natural draft mode. When operated strictly in the natural draft mode, malfunctions of the SXCT fans did not need to be considered because the fans were already assumed to be lost to tornado missiles. The revised analysis considers a bounding electrical bus failure that results in the loss of power to two SXCT fans and a deep well pump. The analysis shows that with the available SXCT fans and risers, cold shutdown conditions can be reached for both units well before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while maintaining the SX supply water temperature less than or equal to the maximum design temperature of 1OO F.

Design analysis of the equipment important to safety located within the MEER and Battery Rooms and the 1B/2B AF Diesel Engine will continue to support safe shutdown with the missile damage acceptance criteria utilized in the TORMIS analysis. The proposed operator actions to mitigate postulated tornado missile damage are simple such that operator errors are judged to not be credible. The potential for MEER ventilation system operation with only the supply fans operating in the outside air mode post-tornado has minimal impact on the control room pressure boundary and the environment in the control room. Thus, the consequences of a tornado event are unchanged and there is no increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The proposed change associated with the evaluation of the impact of tornado missiles does not introduce the possibility of a new accident and no new failure modes are introduced. The acceptance criteria used in TORMIS for missile damage to the 1B/2B AF diesel engine exhaust has a negligible effect on the IB/2B AF engine and the lB/2B

Evaluation No. 6G-J 7-006 50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 4 Page 10 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Change # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMI$ LAR & SUPPORTING DOCSg TRM 3.7.e Tornado Design Basis $XCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown AF pump. The room heat up due the potential loss of MEER ventilation that could result from the impact of tornado missiles does not result in room temperatures that exceed the design limits for equipment important to safety within the rooms. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loss of ventilation in the Battery Rooms will not result in exceeding the hydrogen concentration limits within the rooms or the temperature design limits for the batteries.

The revised UH$ temperature analysis considers a bounding electrical bus failure that results in the loss of power to two SXCT fans and a deep well pump. UFSAR Table 9.2-16, already describes this bounding failure mode. Thus, the change does not introduce an accident different from any previously evaluated in the UF$AR.

The proposed activity adds new procedure instructions and operator actions in the event tornado missile damage affects the MEER exhaust dampers and/or SXCT cooling tower fans. The new procedure instructions and operator actions are included in the operating procedures and are simple tasks. Operator errors are judged to not be credible.

Thus, the change does not introduce an accident different from any previously evaluated in the UfSAR.

The allowed amount of tornado missile damage to the 1B/2B Af Diesel engine exhaust cover plate does not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce a failure result. The acceptance criteria used in TORMIS for missile damage to the 1B/2B Af diesel engine exhaust has a negligible effect on the 1B/2B AF engine and the 1B/2B Af pump.

The potential loss of MEER ventilation due to tornado missile damage to the dampers in the U-line wall does introduce the possibility of a malfunction of an SSC with a different result because the activity does not introduce a failure mode that is not bounded by those described in the MEER ventilation system failure analysis in UFSAR Table 9.4-14. The identified failure mode for this activity is a loss ofthe air flow path due to a missile impact on the exhaust opening. The failure modes and effects analysis for the system in UFSAR Table 9.4-14 includes loss of airflow or flowpath (fan motor continues to operate). A comparison ofthese indicates that the results ofthe failure modes resulting from tornado missile damage to the dampers in the L-line wall is consistent with those presented in the UFSAR.

The proposed change does not introduce the possibility of a malfunction of an SSC with a different result because the activity does not introduce a failure mode that is not bounded by those described in the UHS single failure analysis in UFSAR Table 9.2-16. The revised UHS temperature analysis considers a bounding electrical bus failure that results in the loss of power to two SXCT fans and a deep well pump which bounds the failure modes listed in UfSAR Table 9.2-16. The single failures listed in UFSAR Table 9.2-16 are evaluated for the revised UHS tornado event response in EC 400500. The review determined that the failure modes are bounded by the assumed limiting failure and those presented in the UFSAR.

The revised UHS temperature analysis for safe shutdown from a tornado event shows that with missile damage and the limiting postulated electrical bus failure, cold shutdown conditions can be reached for both units well before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while maintaining the SX supply water temperature less than or equal to the maximum design temperature of 1000 F. The analysis also indicates that the units may be placed on shutdown cooling as early as 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> as long as the rate of cooldown is controlled. The proposed new manual operator actions outside of the control room are limited and can be completed in the time required considering the aggregate effects. Thus, the UHS supports the applicable Branch Technical Position RSB 5-1 design requirements for safe shutdown from a design basis tornado event.

The changes to system parameters were evaluated and the changes do not adversely impact the function of equipment required for safe shutdown from tornado event. Therefore, the proposed activity does not result in a DBLFPB as described in the UFSAR being exceeded or altered.

Evaluation No. 6G-J 7-006 50.59 REVIEW COVER$HEET FORM LS-AA-104-1001 Revision 4 Page 1 1 of 11 Station/Unit(s): Byron/Units 1 and 2 Activity/Document Number: EC 400500, TRM Cliane # 17-004, DRP # 16-030 Revision Number: 0/0/0

Title:

PROCESS TORMIS EAR & SUPPORTING DOCS, TRM 3.7.e Tornado Design Basis SXCT Fans Operating, and TRM 3.7.f Tornado Design Basis SXCT Fans Shutdown The revised UHS temperature analysis for safe shutdown from a tornado event credits SXCT fans and cells that the TORMI$ analysis indicates will be available. This change in the licensing basis to allow credit for some unprotected

$XCT fans and cells based on the results ofthe TORMIS analysis was approved via a NRC Safety Evaluation for Byron Amendment No. 199, dated 8/10/2017. The revised analysis evaluates basin temperature by performing a minute by minute mass and energy balance on the SXCT basin. The specific method used to calculate the maximum basin temperatures in the original analysis was not described the in the UFSAR (fSAR question response) or SER.

The revised analysis uses SXCT performance curves generated using the method described in UFSAR Section 9.2.5.3. 1 1 .2. The specific method used to determine tower performance in the original analysis was not described in the original FSAR question response or SER. The method used for the revised analysis has been previously been reviewed and accepted by the NRC to evaluate tower performance for the Byron UHS design basis event (LOCA coincident with a LOOP on one unit and concurrent orderly shutdown ofthe other unit). The revised UHS temperature analysis assumes that for failed fans that have open riser valves, the SXCT cells operating in the natural draft mode provide 10 percent of the heat removal of a cell with the fan running in high speed. Use of 10% cooling for SXCT operation in the natural draft mode is currently described in UfSAR Section 9.2.5.3.1. 1 .3 and was reviewed and accepted by the NRC in the Safety Evaluation for Byron Station Amendment No. 173 The revised .

UHS temperature analysis utilizes inputs and assumptions consistent with the original UfSAR described analysis.

Thus, the change in to the UHS temperature analysis for safe shutdown from a tornado event does not result in a departure from a method of evaluation described in the UfSAR.

Based on the result ofthe Applicability Review, 50.59 Screening, and 50.59 Evaluation the activity may be implemented per plant procedures.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

Applicability Review 50.59 Screening 50.59 Screening No. 6E-17-047 Rev. 0 50.59 Evaluation 50.59 Evaluation No. 6G-17-006 Rev. 0 See LS-AA-104, Section 5, Documentation, for record retention requirements for this and all other 50.59 forms associated with the Activity.

Evaluation No. 6G-17-007 50.59 REVIEW COVERSHEEtFORM LsA-io4iwii Revision 4 Page 1 of 2 Staffo&Uuk(s): 60.17-007 ActhftylDocument Number PC 621593 RevIsion Number: 0 Tide: Lost Parts Eva] For Reactor Vessel Head Lift Rig Retaining Ring .

NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing die biennial summary report submitted to the NRC in accordance with the requirements of 10 ait 5O.59(dX2).

Descdpdon of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

The activity will evaluate the consequences ofa C-shaped retaining ring on the Reactor Coolantsystem and connected systems ifthe lost part is not recovered. The lost part has dimensions of 1.775 inches by 1.391 inches with a thickness ofO.l50 inches; the ring material is consistent with series 300 stainless steel.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

During an inspection of the Unit 2 Reactor Vessel lifting rig. a retaining ring (1 of 3) was missing during an as-found chect the retaining ring could notbe found. it is unknown how long the part has been not installed. Since the lost part could migrate to the RCS via the refueling cavity, ERAA-2OO6 requires an evaluation to determine the consequences of the lost part on plant equipment.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

There is no effect on the operation of any system. The location of the retaining ring is not known so it is assumed for this evaluation that it could reach the RCS. The potential for damage to plant components is considered les% than minimal based on the retaining rings small size.

Summary of Conclusion for the Activitys 50.59 Review:

(PmvidejusLificaUon for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening. 50.59 Evaluation. or a License Amcndment Request, as applicable, is not required.)

ECs 621593 and 621600 evaluated the effects ofthe retaining ring on RCS. ECCS and fuel components. The lost part will not increase the frequency of occurrence of a stuck control rod. RCP locked rotor. RCP seal failure, RCS decreased flow, or cause a steam generator wbe rupture. Since the likelihood ofan accident is not increased and the components needed to mitigate the consequences to an accident are not more than minimall) impacted. the proposed activity does not result in a more than minimal increase in the consequences ohm accident previously evaluated in the UFSAR.

If the retaining ring is transported to the RCS. the potential to ddmage various applicable components within the RCS and interconnecting systems was evaluated. It was concluded the foreign material does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UPSAR. The intnluetion ofthe retaining ring t th: reactur ma> c.tti* liettu ol tl: fud ci ilJn. hner UPSAR cecti n 4 2 state dte hmit ghen in IOCFRI(%) an I lCFRSU ( sill ant h: evLl\t due In a titill numbs rI rod failures The hkelili .

Of muhipk fuel i xl failmis k unhkel). Tltcieloic. thi.. Itwl. %%sel. RCS. and gnnxtc*1 >taiis and JN11IX)nenh tcrnld nd malfunction in a manna thiferent than ha been pre. iuu%l) e.aluated in the UFS.R 1 hu. the cunsequente u a malfunction are not changed.

The retaining ring is small and would have minimal impact. The lost part does not provide a means to initiate an accident of a different type than any previously evaluated in the UFSAR. The components have been evaluated in ECs 621593 and 621600. and it has been determined that they would continue to perform their design functions and any operational disturbance would be within the plants design and operational basis. Thus, the lost part would not result in an SSC malfunctioning vith a different result than hat iac pre. iousl) e.aluated in the UPSAR.

The three fission product barriers (fuel clad. RCS. containment) ere considered A luel lailure in itself ts ill not etceed or alter a design biis limit foi a fission prnduci harnei (DIII FPB i lutI failuN in uv:ll . dl nnt tcced or alter a design bash limit f.ra fissi n prndui.t barrier (DBLFPB) lvi I I S R %

• tion 4 2. d liiiiit usen in IOCI RlO() and

Evaluation No. 6G-17-007 50.59 REVIEW COVERSEIEET FORM LSA&IO%-IOO!

Revision 4 Page 2 of 2 Station/Unit(s): 6G-17-007__________________________

Activity/Document Number: PC 62 Revision Number: 0 Title; Lost Parts Eval For Reactor Vessel 1-lead Liii Ri% Retaining Rinq________

IOCFR5O.67 will not be exceeded due to a rod Lijiure. Therefore. he lost part do not result in a DBLFPB as described in the UFSAR being exceeded or altered.

The lost part evaluation does not involve any evaluation methodologies used to establish design bases or safety analyses. It is conclutlcd the lost part will not cause more than a minimal impact on plant components antI systems.

Based on this ealuation. the proposed activity can he performed without prior NRC permission per the applicable governing procedures Attachments:

Attach all 5059 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

ci Applicalility Review C 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 5039 Evaluation No. 6G-17-007 Rev. 0 See LSAA 104. Section 5. Documcntation. for recortl retention requirements for this and all other 5059 forms associated with the Activity.