Text

Submittal of 50.59 Evaluation Summary Report
	ML16020A299
Person / Time
Site:	Prairie Island
Issue date:	01/20/2016
From:	Davison K; Northern States Power Co, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-PI-16-002
	Download: ML16020A299 (20)
	v • d • e

(l Xcel Energy JAN 2 0 2016 L-PI-16-002 10 CFR 50.59 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 50.59 Evaluation Summary Report With this letter, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), submits two enclosures. Enclosure 1 contains descriptions and summaries of safety evaluations for changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. Enclosure 2 contains a discussion of changes to regulatory commitments made within our Regulatory Commitment Change Process during the period since the last update.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

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Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota 1717 Wakonade Drive East

Welch, Minnesota 55089-9642 Telephone: 651.388.1121

ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS- DECEMBER 2015 14 Pages Follow

ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS- DECEMBER 2015 Below is a brief description and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out at the Prairie Island Nuclear Generating Plant by Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), without prior Nuclear Regulatory Commission (NRC) approval, pursuant to the requirements of 10 CFR 50.59.

50.59 Evaluation No. 1080, Rev. 1 - Change Licensing Basis for RCS Leakage Detection capability from 1 gpm in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 1 gpm in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (10/24/2013)

Description of Change Revision 1: Revision 1 cancels Revision 0 of this 50.59 evaluation. The RCS leak detection sensitivity and its license bases as discussed in revision 0 of this evaluation was revised with the issuance of Amendment No. 204 to Renewed Facility Operating License No. DRP-42 and Amendment No. 191 to Renewed Facility Operating License No. DRP-60, Prairie Island Nuclear Generating Plant, Unit 1 and Unit 2, respectively.

The amendments are the outcome of License Amendment Request (LAR) titled "Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures From the Licensing Bases Based Upon Application of Leak-Before-Break Methodology". These amendments render the discussion and conclusions of Revision 0 of this evaluation incorrect relative to what is now the current licensing bases.

Revision 0: As the Reactor Coolant System radioactivity levels decreased due to improved fuel performance and reduced general corrosion, the ability to detect a leak based on radio nuclides released into containment decreased. Thus it is no longer practical for the leak detection system to be able to detect a one gpm leak in one hour called for in Prairie Island's Licensing Basis. This evaluation justifies changing the Prairie Island's leak detection system's Licensing Basis from being able to detect an one gpm leak in one hour to being able to detect a one gpm leak in four hours and clarifies that the time to detect a leak, i.e. the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , does not include the time for the leaking fluid to travel from the source to the detection site of the instrument.

Summary of 50.59 Evaluation The applicable criteria for the leak detection system capability listed in Section 5.7 of NUREG-1 061, Volume 3 "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks" calls for a leak detection system with a sensitivity capable of detecting an unidentified leakage rate of one gpm in four hours. Since the proposed change continues to meet this criterion, there is less than a minimal increase in the frequency of occurrence of an accident. The change does not involve any changes to equipment and thus there is no change to the likely-hood or results of a malfunction. The leak detection system only performs a monitoring function and thus does not affect the consequences of, or introduces a different type of, accident or malfunction. The proposed activity does not involve a design basis limit for a fission product barrier or method of evaluation.

Therefore the proposed activity does not require prior NRC approval.

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ENCLOSURE 1 50.59 Evaluation No. 1100, Rev. 0- Unit 2 Replacement Steam Generators- Stress and Fatigue Analysis Report (11/20/2013)

Description of Change The proposed activity being evaluated is the change in the Stress and Fatigue Analysis Report calculations performed for the Unit 2 Replacement Steam Generators. These calculations document the structural analysis and evaluation performed in accordance with the requirements of the design specification and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Division 1, 1995 edition through 1996 Addenda.

Summary of 50.59 Evaluation The structural analysis demonstrates that the Replacement Steam Generators meet the current design basis criteria and limits continue to be met, thus there is no increase in frequency of any accident or creation of an accident of a different type. The methods of analyses that are used are approved by the NRC. The results show that the structural criteria and limits are maintained, thus the safety analyses are unaffected and there is no affect to the consequences of any accident analyses or equipment malfunction.

50.59 Evaluation No. 1104, Rev. 0- Unit 2 Cycle 28 Core Reload Modification (11/26/2013)

Description of Change This activity will replace spent fuel assemblies from the Unit 2 Cycle 27 reactor core with 52 feed fuel assemblies, and rearrange used fuel assemblies of both OFA V+ and 422 V+ design in the Unit 2 Cycle 28 core. This will allow the Unit 2 reactor to produce power at its rated capacity in cycle 28 for approximately 22 months. The Unit 2 Cycle 27 core would only have been able to operate at full capacity for about 3 months if the feed fuel was not added. This evaluation is valid for operation of Unit 2 Cycle 28 in Modes 1 through 6.

Summary of 50.59 Evaluation The UFSAR Chapter 14 evaluations performed by Westinghouse demonstrate that the Prairie Island Unit 2 Cycle 28 reload design and associated COLR do not result in the licensed safety limits for any accident being exceeded. The Cycle 28 design is consistent with the description of the core in the USAR. The core contains 121 fuel assemblies using a 14 X 14 fuel rod array, with 29 control rods in the same locations as described in the UFSAR. The only change from Cycle 27 in the addition of new 422 V+

fuel assemblies and the rearrangement of used fuel assemblies of both the OFA V+

design and the 422V+ design. The change results in an isotopic distribution of the core that changes the core physics parameters. The effect of these changes in the cycle physics parameters on cycle operation and accident analyses have been evaluated using NRC-approved methods discussed in T.S. 5.6.5.

The accident analyses show that no design limits are exceeded during any analyzed transient for the cycle as designed. The cycle does not exceed any fuel burn up limits.

Therefore, the reload modification for Unit 2 Cycle 28 is consistent with Prairie Island Current Licensing Basis and does not need prior NRC approval prior to implementation.

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ENCLOSURE 1 50.59 Evaluation No. 1106, Rev. 1 - Main Steam and Feedwater Break Thrust Reaction Load Methodology (05/07/2014)

Description of Change The proposed activity involves changing an element of the methodology for structurally qualifying the Unit 2 Reactor Coolant loop piping with the Replacement Steam Generators. The element of the methodology involved is the pipe break thrust reaction loads acting on the Replacement Steam Generators by the Primary, Main Steam and Feedwater postulated pipe breaks. These thrust reaction loads were used as inputs in a series of calculations that provide structural qualification to the Reactor Coolant Loop piping, supports and equipment loads. The element of the methodology as described in the Updated Safety Analysis Report includes the initial thrust force being a constant multiplied by the pressure and the area of the postulated break. The prescribed methodology only discusses the initial thrust force and is silent on any subsequent force magnitudes. It is also silent on how this initial thrust force is used in an analysis. The new Reactor Coolant Loop analysis used a computer program that calculated not only the initial thrust force, but subsequent forces as a function of time after the postulated pipe break. The thrust forces were determined as a function of the physical characteristics of the system and piping. The use of the computer program and the thrust forces being calculated as a function of time uses a more modern and accurate representation of the postulated break thereby giving more representative loads.

Summary of 50.59 Evaluation Use of the new element to determine primary, main steam and feedwater pipe break thrust reaction loads does not require prior NRC approval because it calculated conservatively larger reaction load forces when compared to the element used previously.

50.59 Evaluation No. 1107, Rev. 0- Revised LOCA Containment Response Analysis for Unit 1 and Unit 2 (11/26/2013)

Description of Change The purpose of this evaluation is to implement a new calculation as the Analysis of Record for the Unit 1 and Unit 2 containment response to a LOCA (Loss of Coolant Accident) event. This activity is required due to errors found in the current analysis.

When the errors were corrected, additional mass and energy was released into containment. This additional mass and energy is an input into the limiting peak containment pressure and temperature cases for this event.

Analytically the only difference between Prairie Island Unit 1 and Unit 2 has been the steam generators. The current analysis uses inputs from the Unit 1 model 56/19 steam generators to bound the Unit 2 model 51 steam generators. Replacement of the model 51 steam generators to model 56/19 steam generators (same as Unit 1) will not affect the analysis.

Summary of 50.59 Evaluation This activity does not require prior NRC approval as the new analysis of record since the new analysis used the Prairie Island current licensing basis NRC approved 3

ENCLOSURE 1 methodology and the results showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus there is no increase to the consequences of an accident. In addition, this activity does not impact equipment operations, performance and reliability thus there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.

50.59 Evaluation No. 1109, Rev. 0 - EC23288 Determination of Acceptable Baffle-Barrel Bolting (12/20/2013}

Description of Change The proposed activity involves the following changes to the USAR described methodology for reactor vessel internals analysis.

The methodology for evaluation of baffle-barrel bolting and Loss of Coolant Accident (LOCA) hydraulic forces will be changed to the NRC approved methodology in WCAP-15029-P-A.

The methodology for a larger LOCA break opening time will be changed to NRC approved methodology WCAP-14748, Revision 0.

The methodology for combining the LOCA and seismic loads when evaluating the reactor vessel internals will be changed to allow combination using square root sum of the squares (SRSS) per WCAP 16852-P.

The acceptance criteria for reactor vessel internals plastic stress will be changed to allow use of ASME Section Ill Non-mandatory Appendix F, paragraph F-1340, 1998 Edition with Addenda through 2000.

Summary of 50.59 Evaluation Use of the methodology in WCAP-15029-P-A, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions" does not require prior NRC approval, The NRC has found this methodology acceptable to the extent specified and under the limitations delineated in the methodology and in the associated NRC safety evaluation which is attached in the WCAP.

Use of the methodology in WCAP 14748, Revision 0 "Justification for Increasing Postulated Break Opening Times in Westinghouse Pressurizer Water Reactor" does not require prior NRC approval. The NRC has found this methodology acceptable when applied in conjunction with WCAP-15029-P-A.

Use of SRSS to combine LOCA and seismic loads for evaluating reactor vessel internals does not require prior NRC approval. The NRC has found this methodology acceptable as documented in Letter, USNRC to PINGP, Prairie Island Nuclear Generating Plant, Units 1 and 2- Issuance of Amendments Re: Technical Specifications Changes to Allow Use of Westinghouse 0.422-inch OD 14x14 Vantage+ Fuel", Dated July 1, 2009. ML091460809.

Use of ASME Section Ill Non-mandatory Appendix F, paragraph F-1340, 1998 Edition with Addenda through 2000 does not require prior NRC approval. This edition and 4

ENCLOSURE 1 addenda of the code were found acceptable as documented in the letter above. The plastic analysis acceptance criteria in Paragraph F-1340 were not part of the code of construction and is now being adopted.

50.59 Evaluation No. 1110, Rev. 0- EC 23353 "TUBE PLUGGING ON 21 and 22 FCU, COMP MEASURE FOR OPR 1410533" and TMOD EC 23402, "Tube Plugging on 24 FCU, Comp Measure for OPR 1410533" (12/25/2013) (12/23/2013)

Description of Change During 2R28, leaks were identified on the upper and lower south coils of 21 Fan Coil Unit (FCU, 274-011) as well as on the upper east coil of 22 FCU (274-012). These leaks appear to have been caused by the FCU tubes freezing due to the equipment hatch being open during unseasonably cold temperatures. Additional information can be found in ARs 1410533 and 1411076.

The purpose of TMOD EC 23353 is to implement compensatory actions on 21 and 22 FCU. EC 23353 is installing tube plugs on 21 and 22 FCU. Seventeen tubes are being plugged on the 21 FCU and one tube is being plugged on the 22 FCU. No tubes are being plugged on the 23 and 24 FCUs.

50.59 screening 4455 determined that the installation of tube plugs adversely affected the design function of heat removal capabilities and therefore required a 50.59 evaluation.

A compensatory measure for maintaining normal containment temperatures< 105°F are assumed in order to maintain the CFUs heat removal capability during accident conditions.

Summary of 50.59 Evaluation Plugging tubes on the 21 and 22 FCUs does not require prior NRC approval because analysis and testing has demonstrated that even with the plugging, the FCUs will continue to be able to perform their design functions of heat removal, maintaining system pressure boundary and maintaining the containment fission product barrier. In addition, analysis showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus, there is no (less than minimal) increase to the consequences of an accident or malfunction. In addition, since this activity maintains the FCUs design basis functions, there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.

50.59 Evaluation No. 1110, Rev. 1- EC 23353 "TUBE PLUGGING ON 21 and 22 FCU, COMP MEASURE FOR OPR 1410533" and TMOD EC 23402, "Tube Plugging on 24 FCU, Comp Measure for OPR 1410533" (12/25/2013)

Description of Change Revision 1 addresses tube plugging on 24 FCU per EC 23402.

During 2R28, leaks were identified on the upper and lower south coils and lower east coil of 21 Fan Coil Unit (FCU, 274-011 ), the upper east coil of 22 FCU (274-012) and the upper south and upper and lower north coils of 24 FCU (274-014). The 21 and 22 5

ENCLOSURE 1 FCUs were damaged by freezing temperatures during 2R28, however it is not believed that the 24 FCU was subjected to freezing temperatures. The cause of the leak on the 24 FCU has yet to be determined. Additional information can be found in ARs 1410533, 1411076, and 1412237.

The purpose of TMOD EC 23353 is to implement compensatory actions on 21 and 22 FCU by installing tube plugs on 21 and 22 FCU. TMOD EC 23402 will install tube plugs on 24 FCU. Seventeen tubes are being plugged on the 21 FCU, one tube is being plugged on the 22 FCU, and fourteen tubes are being plugged on the 24 FCU. No tubes are being plugged on the 23 FCU.

50.59 screening 4455 determined that the installation of tube plugs adversely affected the design function of the FCUs' heat removal capabilities and therefore required a 50.59 evaluation.

A compensatory measure for maintaining normal containment temperatures < 105°F is assumed in order to maintain the CFUs heat removal capability during accident conditions.

Summary of 50.59 Evaluation Plugging tubes on the 21, 22, and 24 FCUs does not require prior NRC approval because analysis and testing has demonstrated that even with the plugging, the FCUs will continue to be able to perform their design functions of heat removal, maintaining system pressure boundary and maintaining the containment fission product barrier. In addition, analysis showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus, there is no (less than minimal) increase to the consequences of an accident or malfunction. In addition, since this activity maintains the FCUs design basis functions, there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.

50.59 Evaluation No. 1110, Rev. 2- EC 23353 "TUBE PLUGGING ON 21 and 22 FCU, COMP MEASURE FOR OPR 1410533" and TMOD EC 23402, "Tube Plugging on 24 FCU, Comp Measure for OPR 1410533" (12/25/2013) (06/17/2015)

Description of Change During 2R28, leaks were identified on the upper and lower south coils and lower east coil of 21 Fan Coil Unit (FCU, 274-011), the upper east coil of 22 FCU (274-012) and the upper south and upper and lower north coils of 24 FCU (274-014). Additional information can be found inARs 1410533,1411076, and 1412237, and 1477721.

The purpose of TMOD EC 23353 is to implement compensatory actions on 21 and 22 FCU by installing tube plugs on 21 and 22 FCU. TMOD EC 23402 will install tube plugs on 24 FCU.

50.59 screening 4455 determined that the installation of tube plugs adversely affected the design function of the FCUs' heat removal capabilities and therefore required a 50.59 evaluation.

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ENCLOSURE 1 Revision 2 of this evaluation addresses the current amount of tube plugging on the 21 and 22 FCUs as documented in EC 23353 and the current amount of tube plugging on the 24 FCU as documented in EC 23402.

Summary of 50.59 Evaluation Plugging tubes on the 21, 22, and 24 FCUs does not require prior NRC approval because analysis and testing has demonstrated that even with the plugging, the FCUs will continue to be able to perform their design functions of heat removal, maintaining system pressure boundary and maintaining the containment fission product barrier. In addition, analysis showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus, there is no (less than minimal) increase to the consequences of an accident or malfunction. In addition, since this activity maintains the FCUs design basis functions, there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.

50.59 Evaluation No. 1111, Rev. 0 - Install and Remove External Flood Panels with Both Units at Power (03/20/2014)

Description of Change The proposed activity is installing external flooding panels, or angle supports on flood mitigation doors, in the Prairie Island Units 1 and 2 Turbine Buildings (TB) with both units in Mode 1 for the purpose of performing a timing test of the effort. After the timing test has been performed, the plant will be returned to its pre-test (normal) configuration.

This test will be conducted under Work Order 497679 utilizing special test procedure ST FB-INSTALL GP2. The timing test will validate the amount of time it takes to install each of the external flood mitigation panels or angle supports per Attachment J of procedure AB-4, FLOOD. The installation will be performed while both units are in Mode 1 (power operation). This evaluation does not include installing bulkheads over the TB roll-up doors; Doors 44 (Unit 1) and Door 47 (Unit 2). The TB roll-up doors- Doors 44 & 419 11 (Unit 1) and Door 47 (Unit 2)- will be maintained in their normal design (minimum 16 open) position.

Summary of 50.59 Evaluation The proposed activity does not require NRC approval prior to performing the activity because no physical or analytical changes are being made to any Structure, System or Component (SSC) that are accident initiators and no new failure modes are introduced, no physical or analytical changes are being made to any SSCs that are initiators of malfunctions and no new failure modes are introduced. In addition, the plant was originally designed to permit sealing all of the subject doors and acceptable results were attained when the GOTHIC flood model from the TB High Energy Line Break (HELB) flooding record of analysis was run with the subject doors sealed ensuring no SSC would be affected. Finally, no malfunctions beyond those already assumed in any dose analysis are introduced, no new equipment failure modes will be introduced and any existing failure modes will remain unchanged, and there are no methods of evaluation associated with the test.

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ENCLOSURE 1 Therefore, the proposed activity has no effect upon the frequency of occurrence of any accident previously evaluated in the UFSAR, no effect on the likelihood of occurrence of any malfunction previously evaluated in the UFSAR and there is no change in the dose consequences of any accident previously evaluated in the UFSAR. Also, the proposed activity creates no possibility for an accident of a different type than is already analyzed in the UFSAR, there will be no malfunction of any sse important to safety with a different result than any previously evaluated in the UFSAR and it does not result in a design basis limit for a fission product barrier as described in the UFSAR, or any pending submittal, being exceeded or altered.

Finally, there is no departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

50.59 Evaluation No. 1112, Rev. 0- EC17584122 Spent Fuel Pool Heat Exchanger and Component Cooling Water Missile/Seismic Protection (05/07/2014)

Description of Change This 50.59 Evaluation reviews a portion of the modification to the component cooling system. This modification adds redundant air operated valves to the Spent Fuel Pool Heat Exchanger inlet piping and redundant check valves on the same outlet piping in order to isolate the safety related portion from the non-safety related portion of the component cooling water system. The purpose for this modification is to provide protection for the essential portions of the component cooling water system in the event of a tornado missile strike on the unprotected portion of piping, thereby ensuring adequate component cooling water inventory for system operability. This protection is currently being performed by the Auxiliary Building Structure. The automatic isolation on an apparent loss of system integrity will ensure that the system maintains its water inventory. This modification also adds a new automatic action for the Residual Heat Removal heat exchanger isolation motor operated valves to automatically actuate when the new air operated valves close and both component cooling pumps are running. This is to ensure that there is a hydraulic load so that there will be no strong pump-weak pump interactions that could interfere with pump performance. The scope of this evaluation is limited to the potential new failure modes associated with the modification.

The reason for this modification is to allow the use of the Spent Fuel Pool Heat Exchanger which currently outside the area protected by the Auxiliary Building Structure.

Summary of 50.59 Evaluation The proposed activity was determined to not result in more than a minimal increase in the frequency of occurrence or the consequences of an accident, in the likelihood or consequences of occurrence of a malfunction of a system important to safety, in the possibility of an accident of a different type or a malfunction with a different result. It was also found to not result in a design basis limit for a fission product barrier. This activity did not involve any methods of evaluation described in the USAR.

50.59 Evaluation No. 1114, Rev. 0- EC23402 TUBE PLUGGING ON 21/22/24 FCU, COMP MEASURE FOR OPR 1410533, REV. 1 (07/03/2014)

Description of Change 8

ENCLOSURE 1 During 2R28, leaks were identified on the upper and lower south coils and lower east coil of 21 Fan Coil Unit (FCU, 27 4-011 ), the upper east coil of 22 FCU (274-012) and the upper south and upper and lower north coils of24 FCU (274-014). The 21 and 22 FCUs were damaged by freezing temperatures during 2R28, however it is not believed that the 24 FCU was subjected to freezing temperatures. The cause of the leak on the 24 FCU has yet to be determined. Additional information can be found in ARs 1410533, 1411076, and 1412237.

The purpose of TMOD EC 23353 is to implement compensatory actions on 21 and 22 FCU by installing tube plugs on 21 and 22 FCU. TMOD EC 23402 will install tube plugs on 24 FCU. The actual number of tubes plugged remains within the assumptions used in the evaluation of limiting containment internal temperature.

50.59 screening 4455 determined that the installation of tube plugs adversely affected the design function of the FCUs' heat removal capabilities and therefore required a 50.59 evaluation.

Summary of 50.59 Evaluation Plugging tubes on the 21, 22, and 24 FCUs does not require prior NRC approval because analysis and testing has demonstrated that even with the plugging, the FCUs will continue to be able to perform their design functions of heat removal, maintaining system pressure boundary and maintaining the containment fission product barrier. In addition, analysis showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus, there is no (less than minimal) increase to the consequences of an accident or malfunction. In addition, since this activity maintains the FCUs design basis functions, there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.

50.59 Evaluation No. 1115, Rev. 0- Natural Gas Line for Site Administration Building (12/16/2015)

Description of Change The activity being evaluated is the extension of natural gas lines into the Owner Controlled Area to service gas powered appliances in the New Administration Building outside the Protected Area. This evaluation focuses on the effects of a postulated gas line explosion only. A Blast Analysis was performed to support this activity based on the methodology provided in US Nuclear Regulatory Commission Regulatory Guide 1.91, Rev. 2. This analysis determines the minimum safe distance beyond which no adverse effect would occur to plant operation and the blast would not be expected to prevent a safe shutdown. The results of the Blast Analysis show that there are no plant structures, systems or components important to safety within the postulated 1 psi blast wave zone anywhere along the natural gas line. However, it is considered possible that the resulting pressure wave could adversely impact electrical lines around the substation resulting in a Loss Of Offsite Power Event (LOOP). It was determined that the increase in frequency of a LOOP event due to a gas line explosion is less than minimal. The gas service and all effects of a postulated explosion are well outside the Protected Area of 9

ENCLOSURE 1 the plant and therefore do not affect any structures, systems or components inside the Protected Area.

Summary of 50.59 Evaluation By analysis it has been shown that the effects of any postulated explosion due to a rupture of a new gas line in the Owner Controlled Area will not affect any structures, systems or components important to safety inside the Protected Area. It was determined, however, that the postulated explosion could affect electrical systems resulting in a Loss Of Offsite Power (LOOP) event, although the frequency of such an event was found to be less than minimal. The extension of the natural gas lines into the Owner Controlled Area does not result in a more than minimal increase in any accidents or malfunctions previously evaluated in the UFSAR, nor does it have the potential to create a new type of accident or impact fission product barriers. In addition, this activity does not result in departure from any methods of analysis described in the UFSAR.

Therefore, a license amendment is not required to implement this change.

50.59 Evaluation No. 1116, Rev. 0- EC 23519- Unit 1 Cycle 29 Core Reload Modification (10/22/2014)

Description of Change This activity will perform three items:

(1) Replace depleted fuel from the Unit 1 Cycle 28 reactor core with 56 feed (fresh) fuel assemblies and the rearrangement of used fuel assemblies of the 422 Vantage Plus (422V+) design. This will allow the Unit 1 reactor to produce power at its rated capacity in Unit 1 Cycle 29 for approximately 22 months. This activity is required because the fuel in the current core will be depleted to a state that no longer allows for full power operation. This evaluation is valid for operation of Unit 1 Cycle 29 in Modes 1 through 6.

(2) Incorporate a new Analysis of Record (AOR) for the Unit 1 Large Break Lost of Coolant Analysis (LOCA) into the licensing basis. This analysis combines the Unit 1 and Unit 2 analyses into one document now that the Unit 2 steam generators have been replaced. This is a document number change only for Unit 1 (no new analyses were run).

(3) Incorporate a revised Integral Form ZIRLO Cladding Corrosion Model which more accurately reflects the clad surface boiling rate.

Summary of 50.59 Evaluation The UFSAR Chapter 14 safety analysis performed by Westinghouse demonstrate that the Prairie Island Unit 1 Cycle 29 reload design and associated Core Operating Limits Report do not result in the licensed safety limits for any accident being exceeded. The Cycle 29 design is consistent with the description of the core in the USAR. The core contains 121 fuel assemblies using a 14 x 14 fuel rod array, with 29 control rods in the same locations as described in the UFSAR. The only physical change from Cycle 28 is the addition of new 422V+ fuel assemblies and the rearrangement of used fuel assemblies of the 422V+ design. This change results in an isotopic distribution of the core that changes the core physics parameters. The effect of these changes in the cycle 10

ENCLOSURE 1 physics parameters on cycle operation and accident analyses have been evaluated using NRC-approved methods discussed in T.S. 5.6.5.

The accident analyses show that no design limits are exceeded during any analyzed transient for the cycle as designed and the cycle does not exceed any fuel burn up limits. The change described above was evaluated against the eight criteria of 10CFR 50.50(c)(2) and none of the criteria were met. Therefore, the reload modification for Unit 1 Cycle 29 is consistent with Prairie Island's Current Licensing Basis and does not need prior NRC approval prior to implementation.

50.59 Evaluation No. 1118, Rev. 0- EC 24957 Evaluation of Loose Part in the Prairie Island Unit 1 Reactor Vessel (03/18/2015)

Description of Change During Refueling Outage 1R29, a socket head cap screw (SHCS) was unaccounted for and assumed left behind in the Unit 1 Reactor Vessel. Two washers that had been connected to a submarine with this cap screw were found during FOSAR (Foreign Object Search and Retrieval) activities prior to core reload. The cap screw was not known to be missing at the time of the FOSAR and is assumed to be left behind in the reactor vessel below the fuel assemblies.

Engineering Change EC 24957 evaluates this missing socket head cap screw. The basis for this evaluation resides in Westinghouse Letter LTR-R/DA-14-156, Rev. 0, "Evaluation of Loose Part in Prairie Island Unit 1 Reactor Vessel." Structures, systems, and components affected by this change are the structural pieces that make up the envelope below the lower core plate and within the reactor vessel. The activity evaluated in this 50.59 is operation with a cap screw in the RCS.

Summary of 50.59 Evaluation Operation of Cycle 29 with the cap screw will cause less than a minimal increase in the frequency or likelihood of occurrence of RCS pressure boundary ruptures or loss of integrity.

Even if the cap screw were to position itself directly in line with the flow path through the fuel assemblies, the size of the flow restriction is so small that it will have a less than minimal effect on RCS flow through fuel assemblies. There is no increase to the consequences of an accident or malfunction. All design limits continue to be met even with the cap screw in this part of the RCS. No accidents of a different type are created by RCS operation with this extra cap screw. There is no possibility of a malfunction of any components in the RCS with a different result due to the cap screw. Also, this activity does not affect any methods of evaluation as described in the USAR.

Based on this 50.59 evaluation, prior NRC approval is not required for plant operation with this loose part inside the RCS.

The cap screw does not adversely affect any conditions described or analyzed in the USAR or other relevant documents.

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ENCLOSURE 1 50.59 Evaluation No. 1119, Rev. 1 - EC 22499 Instrument Air Compressor Replacement (05/20/2015)

Description of Change The activity being evaluated is a potential software common cause failure of the new digital controls for the new Instrument Air Compressors being installed at Prairie Island.

The common cause failure is a non-specific type failure that is assumed because it is not possible to rule out this type of failure due to the proprietary nature of the vendor-supplied software. 50.59 Screening #4671 was written to screen out the non-adverse aspects of the Instrument Air Compressor replacement modification. This evaluation focuses on the software common cause failure only. These same compressors with their associated controls and software have been used in Prairie Island's Station Air system successfully for the last four years. This 50.59 Evaluation evaluates the effects of a postulated common cause failure on the Instrument Air system taking into consideration the system's importance to safety.

Summary of 50.59 Evaluation This evaluation has determined that NRC approval is not required to make the described change to the facility. The evaluation shows that there is no more than a minimal increase in the frequency of occurrence of any accidents or malfunctions previously evaluated in the USAR, nor are there any new types of accidents or malfunctions with different results introduced as a result of the proposed change. Also, since the Instrument Air system does not directly contribute to accident response and mitigation, there are no impacts on the consequences of any accident or malfu,nction previously evaluated in the USAR. This activity does not involve any DBLFBs or methods of evaluation described in the USAR. Therefore, NRC approval is not required for the proposed change.

50.59 Evaluation No. 1120, Rev. 0- EC23207, AVR Digital System Upgrade and Software Common Mode Failure (05/06/2015)

Description of Change The activity being evaluated is a potential software common cause failure of the new digital controls and combination of multiple functions into one digital device for the replacement Automatic Voltage Regulator (AVR) for the Main Electric Generator. The common cause failure is a non-specific type failure that is assumed because failure of the AVR module has been identified as a possible malfunction which would result in a generator trip. The previous system had standalone components that performed various monitoring and control functions of the AVR system; several of these components have now been combined under one digital system, which may increase the likelihood of a malfunction or accident.

Summary of 50.59 Evaluation This evaluation determines that NRC approval is not required to make the described change to the facility. The evaluation shows that there is no more than a minimal increase in the frequency of occurrence of any accidents or malfunctions previously evaluated in the USAR nor are there any new types of accidents of malfunctions with 12

ENCLOSURE 1 different results introduced as a result of this proposed change. Also, since the generator does not directly contribute to accident response and mitigation, there are no impacts on the consequences of any accident or malfunction previously evaluated in the USAR. This activity does not involve any DBLFBs or methods of evaluation described in the USAR. Therefore, prior NRC approval is not required for the proposed change.

50.59 Evaluation No. 1120, Rev. 1 - EC23207, AVR Digital System Upgrade and Software Common Mode Failure (08/21/2015)

Description of Change The activity being evaluated is a potential software common cause failure of the new digital controls and combination of multiple functions into one digital device for the replacement Automatic Voltage Regulator (AVR) for the Main Electric Generator. The common cause failure is a non-specific type failure that is assumed because failure of the AVR module has been identified as a possible malfunction which would result in a generator trip. The previous system had standalone components that performed various monitoring and control functions of the AVR system; several of these components have now been combined under one digital system, which may increase the likelihood of a malfunction or accident.

Summary of 50.59 Evaluation This evaluation determines that NRC approval is not required to make the described change to the facility. The evaluation shows that there is no more than a minimal increase in the frequency of occurrence of any accidents or malfunctions previously evaluated in the USAR nor are there any new types of accidents of malfunctions with different results introduced as a result of this proposed change. Also, since the generator does not directly contribute to accident response and mitigation, there are no impacts on the consequences of any accident or malfunction previously evaluated in the USAR. This activity does not involve any DBLFBs or methods of evaluation described in the USAR. Therefore, prior NRC approval is not required for the proposed change.

50.59 Evaluation No. 1123, Rev. 0- EC26230, Unit 2 Cycle 29 Core Reload Modification (11/04/2015)

Description of Change This activity will perform two items:

(1). replace depleted fuel from the Unit 2 Cycle 28 reactor core with 56 feed (fresh) fuel assemblies and the rearrangement .of used fuel assemblies of the 422 Vantage Plus (422V+) design. This will allow the Unit 2 reactor to produce power at its rated capacity in Unit 2 Cycle 29 for approximately 22 months. This activity is required because the fuel in the current core will be depleted to a state that no longer allows for full power operation. This evaluation is valid for operation of Unit 2 Cycle 29 in Modes 1 through 6.

(2). incorporate a revised Integral Form ZIRLO Cladding Corrosion Model which more accurately reflects the clad surface boiling rate.

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ENCLOSURE 1 Summary of 50.59 Evaluation The UFSAR Chapter 14 safety analysis performed by Westinghouse demonstrate that the Prairie Island Unit 2 Cycle 29 reload design and associated Core Operating Limits Report do not result in the licensed safety limits for any accident being exceeded. The Cycle 29 design is consistent with the description of the core in the USAR. The core contains 121 fuel assemblies using a 14 x 14 fuel rod array, with 29 control rods in the same locations as described in the UFSAR. The only physical change from Cycle 28 is the addition of new 422V+ fuel assemblies and the rearrangement of used fuel assemblies of the 422V+ design. This change results in an isotopic distribution of the core that changes the core physics parameters. The effect of these changes in the cycle physics parameters on cycle operation and accident analyses have been evaluated using NRC-approved methods discussed in T.S. 5.6.5.

The accident analyses show that no design limits are exceeded during any analyzed transient for the cycle as designed and the cycle does not exceed any fuel burn up limits. The change described above was evaluated against the eight criteria of 10CFR 50.50(c)(2) and none of the criteria were met. Therefore, the reload modification for Unit 2 Cycle 29 is consistent with Prairie Island's Current Licensing Basis and does not need prior NRC approval prior to implementation.

Unit 2 Cycle 29 will be the first Unit 2 cycle that employs a new NRC approved Integral Form ZIRLO Cladding Corrosion model, which introduces an instantaneous boiling rate term on the clad surface temperature and more accurately reflects the temperature profile in the boiling region required to predict the measured oxides. Its use does not constitute a change to a method of evaluation, but will require a USAR change to update the corrosion limits in Sections 3.2.1 and 3.4.3.1.

50.59 Evaluation No. 1124, Rev. 0 - EC26230, Incorporate WCAP-12472-P-A Addendum 1-A "BEACON' Core Monitoring and Operations Support System" into PI's Licensing Bases (10/27/2015)

Description of Change The purpose of this evaluation is to incorporate the method of evaluation described in WCAP-12472-P-A Addendum 1-A "BEACON' Core Monitoring and Operations Support System" into Prairie Island's Licensing Bases. This activity is required because Prairie Island has been using the methodology outlined in this WCAP addendum, but has not listed it in any licensing bases documents.

Summary of 50.59 Evaluation This activity does not require prior NRC approval as the BEACON' methodology discussed in WCAP-12472-P-A Addendum 1-A, "BEACON' Core Monitoring and Operations Support System" provides essentially the same results as the currently licensed methodology and the new methodology has been previously approved by the NRC for application at 2-loop Pressurized Water Reactors (PWRs) such as Prairie Island Units 1 and 2.

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ENCLOSURE 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CHANGES TO REGULATORY COMMITMENTS 3 Pages Follow

ENCLOSURE 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CHANGES TO REGULATORY COMMITMENTS Regulatory Commitment Change P880008A- Change commitment wording to commitment in 1992 in response to NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems" (TAC NOS 69673 and 69674).

Commitment Source Document

NRC Response to NSPM, " NRC BULLETIN 88-08, "Thermal Stresses In Piping Connected to Reactor Coolant Systems" (TAC NOS. 69673 AND 69674),"' dated November 4, 1991, Accession No. ML082880483.

NSPM letter to NRC, "Response to NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems," dated January 31, 1992, Accession No.

9202040411.

Purpose for Original Commitment The purpose of the commitment is to response to NRC Bulletin 88-08, Action 3 and Supplement

4. Action 3 states "Plan and implement a program to provide continuing assurance that unisolable sections of all piping connected to the RCS will not be subjected to combined cyclic and static thermal and other stresses that could cause fatigue failure during the remaining life of the unit This assurance may be provided by (1) redesigning and modifying these section of piping to withstand combined stresses caused by various loads including temporal and spatial distributions of temperature resulting from leakage across valve seats, (2) instrumenting this piping to detect adverse temperature distributions and establishing appropriate limits on temperature distributions, or (3) providing means for ensuring that pressure upstream from block valves which might leak is monitored and does not exceed RCS pressure.

Original Commitment Wording "While our 1988 response to Bulletin 88-08 did not commit to continuous temperature monitoring, we have since then implemented temperature monitoring for both Prairie Island units. We have determined that this monitoring complies with the guidelines given in the referenced "Evaluation Criteria for Responses to NRC Bulletin.88-08, Action 3 and Supplement 4."

Revised Commitment Wording "Prairie Island commits to implementation of a program to provide continuing assurance that unisolable sections of all piping connected to the RCS will not be subjected to combined cyclic and static thermal and other stresses that could cause fatigue failure during the remaining life of the unit This program is aligned with industry NEI 03-08, "Guideline for the Management of Materials 1

ENCLOSURE 2 Issues", "Needed" elements for management of thermal fatigue in normally stagnant non-isolable reactor coolant system branch lines."

Justification for Change The original commitment was for temperature monitoring of auxiliary spray lines.

Industry initiatives subsequent to Bulletin 88-08 have determined that the auxiliary spray piping is not susceptible to temperature distributions or oscillations which could cause unacceptable thermal stresses. The purpose of the monitoring, which is to ensure unisolable RCS branch piping does not experience thermal fatigue cracking, is retained by meeting the NEI 03-08 "needed" requirements associated with MRP-146.

Since publication of Bulletin 88-08, the industry has continued evaluation of the phenomena of thermal stratification and cyclic fatigue in RCS unisolable branch piping. In June 2005, EPRI issued NEI 03-08 "Needed" Guidance associated with MRP-146, Rev. 0, "Management of Thermal Fatigue in Normally Stagnant Non-lsolable Reactor Coolant System Branch Lines". In Jan 2009, EPRI issued NEI 03-08 "Needed" guidance for MRP-146S, Rev. 0, Management of Thermal Fatigue in Normally Stagnant Non-lsolable Reactor Coolant System Branch Lines - Supplemental Guidance (1018330). In June 2011, EPRI issued NEI 03-08 "Needed" guidance for MRP-146 Rev. 1, Management of Thermal Fatigue in Normally Stagnant Non-lsolable Reactor Coolant System Branch Lines (1 022564). In May 2015, EPRI issued letter MRP 2015-019 containing NEI 03-08 "Needed" Interim Guidance requirements for management of thermal fatigue supplementing the existing guidelines for normally isolated RCS branch lines. These documents describe the evaluation and inspections/monitoring requirements for lines which may be susceptible to the phenomena described by Bulletin 88-08.

Regulatory Commitment Change N00774- Change to commitment meets the intent of GL 89-13 for performance testing large heat exchangers to verify heat transfer capabilities.

Commitment Source Document

NSPM letter to NRC, "Response to Generic Letter 89.13 Service Water Systems Problems Affecting Safety Related Equipment," dated January 29, 1990, Accession No. 9002060101 Purpose for Original Commitment Meets the intent of GL 89-13 for performance testing large heat exchangers to verify heat transfer capabilities.

Original Commitment Wording "In response to Generic Letter 89-13, implement a periodic program to retest for degraded performance of the large Cooling Water safety-related heat exchangers. The minimum retest frequency for CFCU testing is 5 years +/- 20% frequency."

2

ENCLOSURE 2 Revised Commitment Wording "In response to Generic Letter 89-13, implement a periodic program to retest for degraded performance of the large Cooling Water safety-related heat exchangers. The minimum retest frequency as listed in Generic Letter 89-13 is 5 years+/- 20% frequency. The first performance test may be deferred for replacement heat exchangers designed for equal or better heat transfer performance than the heat exchangers to be replaced; testing will resume at the minimum test frequency after the heat exchanger has been replaced."

Justification for Change Degradation mechanisms described in GL 89-13 would not be found in new heat exchangers since replaced heat exchangers are free from silt and biofouling. After the replaced heat exchanger has accumulated time in service, performance testing will resume at the previously defined frequency. Previous testing has determined that the minimum test frequency of 5 years for the Containment Fan Coil Units (CFCUs) is appropriate since no significant fouling has been discovered. The planned replacement CFCU, design has an equivalent or better heat transfer as determined by Super Radiator Coils Performance Topical Report SRC-TR-6027910-1.

3

(l Xcel Energy JAN 2 0 2016 L-PI-16-002 10 CFR 50.59 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 50.59 Evaluation Summary Report With this letter, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), submits two enclosures. Enclosure 1 contains descriptions and summaries of safety evaluations for changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. Enclosure 2 contains a discussion of changes to regulatory commitments made within our Regulatory Commitment Change Process during the period since the last update.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

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Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota 1717 Wakonade Drive East

Welch, Minnesota 55089-9642 Telephone: 651.388.1121