L-MT-22-021, 0 to Updated Safety Analysis Report, Appendix K, Revision 37, Renewed Operating License-USAR Supplement

0 to Updated Safety Analysis Report, Appendix K, Revision 37, Renewed Operating License-USAR Supplement

ML23006A157
Person / Time
Site:	Monticello
Issue date:	12/19/2022
From:	Xcel Energy, Northern States Power Company, Minnesota
To:	Office of Nuclear Reactor Regulation
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ML23006A159	List: L-MT-22-021, 0 to Updated Safety Analysis Report, Appendix K, Revision 37, Renewed Operating License-USAR Supplement ML23006A127 ML23006A128 ML23006A129 ML23006A130 ML23006A132 ML23006A133 ML23006A134 ML23006A135 ML23006A136 ML23006A137 ML23006A138 ML23006A139 ML23006A140 ML23006A141 ML23006A142 ML23006A143 ML23006A144 ML23006A145 ML23006A146 ML23006A147 ML23006A148 ML23006A149 ML23006A150 ML23006A151 ML23006A152 ML23006A153 ML23006A155 ML23006A156
References
L-MT-22-021
Download: ML23006A157 (1)

Text

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 1 of 84 TABLE OF CONTENTS K1 Appendix K Introduction ........................................................................................... 3 K2 Programs that Manage the Effects of Aging ............................................................ 3 K2.1 Aging Management Programs............................................................................... 4 K2.1.1 10 CFR 50, Appendix J .................................................................................... 4 K2.1.2 ASME Section XI In-Service Inspection, Subsections IWB, IWC, and IWD ..... 4 K2.1.3 ASME Section XI, Subsection IWF .................................................................. 5 K2.1.4 Bolting Integrity ................................................................................................ 5 K2.1.5 Buried Piping & Tanks Inspection .................................................................... 6 K2.1.6 Bus Duct Inspection Program .......................................................................... 8 K2.1.7 BWR Control Rod Drive Return Line Nozzle .................................................... 9 K2.1.8 BWR Feedwater Nozzle................................................................................... 9 K2.1.9 BWR Penetrations ......................................................................................... 10 K2.1.10 BWR Stress Corrosion Cracking .................................................................... 10 K2.1.11 BWR Vessel ID Attachment Welds ................................................................ 11 K2.1.12 BWR Vessel Internals .................................................................................... 11 K2.1.13 Closed-Cycle Cooling Water System ............................................................. 12 K2.1.14 Compressed Air Monitoring ........................................................................... 14 K2.1.15 Electrical Cables & Connections Not Subject To 10 CFR 50.49 Environmental Qualification Requirements Program ............................................................. 14 K2.1.16 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrument Circuits ..................................................... 16 K2.1.17 Fire Protection ............................................................................................... 17 K2.1.18 Fire Water System ......................................................................................... 17 K2.1.19 Flow-Accelerated Corrosion ........................................................................... 18 K2.1.20 Fuel Oil Chemistry ......................................................................................... 19 K2.1.21 Inaccessible Medium Voltage (2kV to 34.5kV) Cables Not Subject to 10 CFR 50.49 EQ Requirements .......................................................................................... 19 K2.1.22 Inspection of Overhead Heavy Load & Light Load (Related to Refueling) Handling Systems ......................................................................................................... 20 K2.1.23 One-Time Inspection...................................................................................... 21 K2.1.24 Open-Cycle Cooling Water System ............................................................... 22 K2.1.25 Plant Chemistry Program ............................................................................... 22 K2.1.26 Primary Containment In-Service Inspection Program .................................... 23

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 2 of 84 K2.1.27 Protective Coating Monitoring & Maintenance Program ................................ 23 K2.1.28 Reactor Head Closure Studs ......................................................................... 24 K2.1.29 Reactor Vessel Surveillance .......................................................................... 24 K2.1.30 Selective Leaching of Materials ..................................................................... 25 K2.1.31 Structures Monitoring Program ...................................................................... 25 K2.1.32 System Condition Monitoring Program .......................................................... 27 K2.1.33 Thermal Aging & Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) ................................................................................................. 28 K2.1.34 Electrical Cable Connections Not subject to 10 CFR 50.49 Environmental Qualification Requirements ............................................................................ 29 K2.1.35 Outdoor and Large Atmospheric Metallic Storage Tanks (OLAMST) ............ 30 K3 Evaluation of Time-Limited Aging Analyses ........................................................... 31 K3.1 Neutron Embrittlement of the Reactor Pressure Vessel and Internals ................ 32 K3.2 Metal Fatigue of the RPV and Internals, and Reactor Coolant Pressure Boundary Piping and Components ...................................................................................... 50 K3.3 ASME Section III Class 1 Reactor Coolant Pressure Boundary (RCPB) Piping Fatigue Analysis ............................................................................................................... 55 K3.4 RCPB Section III Class 2 AND 3, ASA B31.1 AND, USAS B31.1 Piping and Components ........................................................................................................ 57 K3.5 Irradiation Assisted Stress Corrosion Cracking ................................................... 58 K3.6 Stress Relaxation of Rim Holddown Bolts ........................................................... 59 K3.7 Effects of Reactor Coolant Environment ............................................................. 59 K3.8 Fatigue Analyses of the Primary Containment, Attached Piping, and Components62 K3.9 Environmental Qualification of Electrical Equipment (EQ)................................... 68 K3.10 Reactor Building Crane Load Cycles .................................................................. 71 K3.11 Fatigue Analyses of HPCI & RCIC Turbine Exhaust Penetrations ...................... 72 K4 TLAA Supporting Activities .................................................................................... 74 K4.1 Environmental Qualification (EQ) of Electrical Components ............................... 74 K4.2 Metal Fatigue of Reactor Coolant Pressure Boundary ........................................ 74 K4.3 Exemptions ......................................................................................................... 74 K.5 Commitments ......................................................................................................... 75 References ....................................................................................................................... 82

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 3 of 84 K1 Appendix K Introduction This appendix was originally included as Appendix A in the MNGP License Renewal Application (LRA). It has been revised from the original version as a result of the License Renewal audit and review process. The revised Appendix A includes the commitments in the individual program descriptions described in Section A2.1, Aging Management Programs, and adds a new Section A.5, Commitments, that provides a table summarizing the actions committed to by Nuclear Management Company LLC, (NMC) in the License Renewal Application for the MNGP. The only difference between the revised and updated LRA Appendix A and this appendix is the addition of this paragraph. The appendix designation change from A to K to correspond with the USAR appendix designations, and the changes from NMC to NSPM reflecting the operating license transfer from NMC to the NSPM (Reference 21).

The application for a renewed operating license is required by 10 CFR 54.21(d) to include an Updated Safety Analysis Report (USAR) Supplement. This appendix provides the required supplement for the Monticello Nuclear Generating Plant (MNGP)

USAR.

Section K2 of this appendix contains a summary description of the programs for managing the effects of aging during the period of extended operation. Section K3 contains a summary of the evaluation of time-limited aging analyses (TLAAs) for the period of extended operation. Section K4 contains summaries of TLAA supporting activities. Section K5 contains the table of commitments related to License Renewal Aging Management for the Monticello Nuclear Generating Plant.

K2 Programs that Manage the Effects of Aging This section provides summaries of the programs and activities, in alphabetical order, credited for managing the effects of aging. These aging management programs may not exist as discrete programs at MNGP. In many cases they exist as a compilation of various implementing documents that, when taken as a whole, satisfy the intent of NUREG-1800 and/or NUREG-1801 attributes.

The MNGP Quality Assurance Program (USAR Appendix C) implements the requirements of 10 CFR 50, Appendix B, and is consistent with the summary in Appendix A.2 of NUREG-1800, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants, published July 2001. The elements of corrective action, confirmation process, and administrative controls in the Quality Assurance Program are applicable to both safety related and non-safety related systems, structures, and components that are subject to an aging management review.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 4 of 84 K2.1 Aging Management Programs K2.1.1 10 CFR 50, Appendix J Program Description The MNGP 10 CFR 50, Appendix J Program specifies pneumatic pressure tests and visual examinations to verify the structural and leak tight integrity of the primary containment. An overall (Type A) pressure test assesses the capacity of the containment to retain design basis accident pressure. This test also measures total leakage through the containment pressure-retaining boundary. Local (Type B & C) tests measure leakage through individual penetration isolation barriers. These barriers are maintained as required to keep overall and local leakage under Technical Specification and plant administrative limits.

Tests are performed at intervals determined by the risk and performance factors applicable to each tested item in accordance with governing regulations and standards. This risk and performance based approach to testing provides reasonable assurance that developing leakage is detected and corrected well before it reaches a magnitude that could compromise containment function.

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test, and during at least three other outages before the next Type A test, if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE required examinations. The Primary Containment Inservice Inspection Aging Management Program (EWI-11.01.30) performs general visual examinations that assess overall structural condition once during each inservice inspection period.

K2.1.2 ASME Section XI In-Service Inspection, Subsections IWB, IWC, and IWD Program Description The MNGP ASME Section XI In-Service Inspection, Subsections IWB, IWC, and IWD Program is part of the MNGP ASME Section XI In-Service Inspection Program. The program is also supplemented by implementing the guidelines of the Boiling Water Reactor Integrity Project (BWRVIP) program documents. The program provides for condition monitoring of Class 1, 2, and 3 pressure-retaining components and their integral attachments.

Piping is inspected in accordance with NRC approved alternatives using Risk-Informed (RI) and Risk-Informed Safety-Based (RIS_B) methodologies, which are largely based on Electric Power Research Institute (EPRI) "Topical Report TR-112657, Rev. B-A, Revised Risk Informed Inservice Inspection Evaluation Procedure" and related ASME Section XI Code Cases.

The approved alternatives are included in the applicable ISI Interval Plan.

The program is updated periodically as required by 10 CFR 50.55a.

The Plant Chemistry Program augments this program where applicable.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 5 of 84 Augmented volumetric examinations of welds are performed on NPS 2 inch through less than NPS 4 inch stainless steel Class 1 small bore piping butt welds.

K2.1.3 ASME Section XI, Subsection IWF Program Description The MNGP ASME Section XI, Subsection IWF Program is part of the MNGP ASME Section XI In-Service Inspection Program. The ASME Section XI, Subsection IWF Program is performed in accordance with the approved ASME Section XI Edition /

Addenda and 10 CFR 50.55a and provides for condition monitoring of Class 1, 2, 3, and MC component supports. Component supports are selected for inspection in accordance with the ASME code classification. The quantity of component supports selected for examination is increased as a result of discovered support deficiencies.

Visual inspection is the primary method for identifying deficiencies.

The program is updated periodically as required by 10 CFR 50.55a.

Commitments Prior to the period of extended operation, the MNGP ASME Section Xl, Subsection IWF Program was enhanced to provide inspections of Class MC component supports consistent with NUREG-1801, Chapter III Section B1.3.

K2.1.4 Bolting Integrity Program Description The Bolting Integrity Program manages the aging affects associated with bolting in the scope of license renewal through periodic inspection, material selection, thread lubricant control, assembly and torque requirements, and repair and replacement requirements. These activities are based on the applicable requirements of ASME Section XI and plant operating experience and includes consideration of the guidance contained in NUREG-1339, Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants, EPRI NP-5769, Degradation and Failure of Bolting in Nuclear power Plants, EPRI TR-104213, Bolted Joint Maintenance & Application Guide, and EPRI NP-5067 Volumes 1 and 2, Good Bolting Practices.

The program credits other MNGP Aging Management Programs for the inspection of installed bolts. These other programs are:

• 10 CFR 50, Appendix J,

• ASME Section XI In-Service Inspection, Subsections IWB, IWC and IWD,

• Primary Containment In-Service Inspection,

• Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems,

• ASME Section XI, Subsection IWF,

• Buried Piping and Tanks Inspection,

• Bus Duct Inspection,

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 6 of 84

• BWR Vessel Internals,

• Reactor Head Closure Studs Monitoring,

• System Condition Monitoring, and 604000000841

• Structures Monitoring.

• Outdoor and Large Atmospheric Metallic Storage Tanks (OLAMST)

Commitments Prior to the period of extended operation, the guidance for performing visual bolting inspections contained in EPRI TR-104213, Bolted Joint Maintenance & Application Guide, and the Good Bolting Practices Handbook (EPRI NP-5067 Volumes 1 and 2) has been included in the Bus Duct Inspection Program, Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program, Structures Monitoring Program and the System Condition Monitoring Program.

As a result of reviews for newly identified SSCs in scope of license renewal aging management in accordance with 10 CFR 54.37(b), this program's scope has been 604000000815 expanded to include components within the boundary of the Alternate Source Term (AST) and the following component group(s) have had "Plateout and holdup of radioactive materials" added as an 'intended function':

• Fasteners/bolting K2.1.5 Buried Piping & Tanks Inspection Program Description The buried piping and tanks inspection program consists of preventive and condition monitoring measures to manage the aging effects for buried piping, conduit and tanks in scope for license renewal. Buried components in scope for license renewal include carbon steel piping, bolting, conduit and tanks (loss of material due to general, crevice, galvanic, MIC and pitting corrosion), stainless steel (loss of material due to crevice, MIC and pitting corrosion) and cast iron piping (loss of material due to general, crevice, galvanic, MIC and pitting corrosion and selective leaching).

Preventive measures consist of protective coatings and/or wraps on buried components. Condition monitoring consists of periodic inspections of buried components.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 7 of 84 In addition, buried components are not routinely uncovered during maintenance activities. Therefore, other system monitoring and functional testing activities are relied upon to provide effective degradation aging management of buried piping and tanks. Some of these activities are neither preventive nor mitigative in nature, but they do provide indication of a leak. However, the potential problem is detected at an early stage, i.e. small leak, such that repairs can be made prior to loss of component intended function.

Commitments Prior to the period of extended operation:

1) The Buried Piping and Tanks Inspection Program has updated the implementing procedures to include inspections of buried components when they are uncovered.

2) The Diesel Fuel Oil Storage Tank, T-44, internal inspection has been added to the list of scheduled inspections in the Buried Piping and Tanks Inspection Program.

3) The Buried Piping and Tanks Inspection Program has been revised to include a provision that if evaluations of pipe wall thickness show a susceptibility to corrosion, further evaluation as to the extent of susceptibility will be performed.

4) The Buried Piping and Tanks Inspection Program has been revised to specify a 10-year buried pipe inspection frequency.

5) The Buried Piping and Tanks Inspection Program has been revised to specify a 10-year inspection frequency for Diesel Fuel Oil Storage Tank T-44.

6) The Buried Piping and Tanks Inspection Program has been revised to include a review of previous buried piping issues to determine possible susceptible locations.

7) As a result of reviews for newly identified SSCs in scope of license renewal aging management in accordance with 10 CFR 54.37(b), this program's scope has been 604000000815 expanded to include components within the boundary of the Alternate Source Term (AST) and the following component group(s) have had "Plateout and holdup of radioactive materials" added as an 'intended function':

• Piping and Fittings

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 8 of 84 K2.1.6 Bus Duct Inspection Program Program Description The purpose of this aging management program is to demonstrate, for in-scope non-segregated bus ducts, that the aging effects caused by ingress of moisture or contaminants (dust and debris), insulation degradation caused by heat or radiation in the presence of oxygen, and bolt relaxation caused by thermal cycling will be adequately managed so that there is reasonable assurance that the non-segregated bus ducts will perform their intended function in accordance with the current licensing basis during the period of extended operation. The intended function of non-segregated bus ducts is to provide electrical connections to specified sections of an electrical circuit to deliver voltage, current or signals.

Industry operating experience indicates that the failure of bus ducts is caused by the cracking of bus bar insulation (bus sleeving) combined with the accumulation of moisture or debris. Cracked insulation in the presence of moisture or debris provides phase-to-phase or phase-to-ground electrical tracking paths, which can result in catastrophic failure of the buses.

Bus ducts exposed to appreciable ohmic heating during operation may experience loosening of bolted connections because of the repeated cycling of connected loads.

This phenomenon can occur in heavily loaded circuits, i.e., those exposed to appreciable ohmic heating. Sandia 96-0344 identified instances of bolted connection loosening at several plants due to thermal cycling. NRC Information Notice 2000-14 identified torque relaxation of splice plate connecting bolts as one potential cause of a bus duct fault.

The primary objective of the aging management program is to provide an inspection of bus ducts. Non-segregated bus duct insulation aging degradation from ingress of moisture or contaminants (dust and debris), or heat or radiation in the presence of oxygen causes insulation surface anomalies. In managing this aspect of the aging management program, visual inspection of interior portions of bus ducts will be performed to identify aging degradation of insulating and metallic components and water/debris intrusion. The external portions of bus ducts and structural supports will be inspected in accordance with a plant specific structural monitoring program.

Additionally, bus ducts exposed to appreciable ohmic heating during operation may experience loosening of bolted connections. In managing this aspect of the aging management program, bolted connections at sample sections of the buses in the metal enclosed bus (MEB) will be checked for loose connections by using thermography or by measuring connection resistance using a low range ohmmeter. As an alternative to thermography or measuring connection resistance of bolted connections, for the accessible bolted connections that are covered with heat shrink tape, sleeving, insulating boots, and other materials, MNGP will use visual inspection of insulation material to detect surface anomalies, such as discoloration, cracking, chipping or surface contamination.

The purpose of the aging management program is to provide reasonable assurance that the intended functions of nonsegregated bus ducts that are not subject to the environmental qualification requirements of 10 CFR 50.49 and are exposed to adverse localized environments caused by the ingress of moisture, contaminants (dust and debris), insulation degradation caused by heat or radiation in the presence

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 9 of 84 of oxygen, and bolt relaxation caused by thermal cycling will be maintained consistent with the current licensing basis through the period of extended operation. This program considers the technical information provided in Information Notice No. 89-64.

Commitments Prior to the period of extended operation, the Bus Duct Inspection Program has been implemented consistent with the appropriate ten elements described in Appendix A of NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants.

K2.1.7 BWR Control Rod Drive Return Line Nozzle Program Description The MNGP BWR Control Rod Drive Return Line Nozzle Program is part of the MNGP ASME Section XI In-Service Inspection Program. The BWR Control Rod Drive Return Line Nozzle Program is in accordance with the approved ASME Section XI Edition /

Addenda and provides for condition monitoring of the BWR Control Rod Drive Return Line (CRDRL) nozzle.

In 1977 the CRDRL nozzle safe end was removed and the CRDRL nozzle was capped. In 1986 the CRDRL nozzle was modified again by removing the portion of the existing weld butter layer susceptible to IGSCC, by re-cladding the weld prep area with corrosion resistant cladding, and by installing a new nozzle cap of non-IGSCC susceptible stainless steel. As a result of capping the CRDRL nozzle, the NUREG-0619 augmented examinations are no longer required. Not performing the NUREG-0619 augmented examinations is considered a NUREG-1801 XI.M6 program exception.

The program is updated periodically as required by 10 CFR 50.55a.

K2.1.8 BWR Feedwater Nozzle Program Description The MNGP BWR Feedwater Nozzle Program is part of the MNGP ASME Section XI In-Service Inspection Program. The BWR Feedwater Nozzle Program is in accordance with the approved ASME Section XI Edition / Addenda with Appendix VIII.

The program provides for condition monitoring of the BWR feedwater nozzles. The BWR feedwater nozzles were all repaired in 1977 and the safe ends were all replaced in 1981 with a tuning fork design with a welded-in thermal sleeve.

The BWR Feedwater Nozzle Program has been enhanced to include the recommended inspection regions specified in General Electric (GE)

NE-523-A71-0594-A, Revision 1, Alternate BWR Feedwater Nozzle Inspection Requirement.

The Program is updated periodically as required by 10 CFR 50.55a.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 10 of 84 Commitments Prior to the period of extended operation, the BWR Feedwater Nozzle Program has been enhanced so:

1) Regions being inspected are consistent with the recommendations of GE NE-523-A71-0594-A, Revision 1.

K2.1.9 BWR Penetrations Program Description The MNGP BWR Penetrations Program is part of the MNGP ASME Section XI In-Service Inspection Program. The BWR Penetrations Program is in accordance with the approved ASME Section XI Edition / Addenda and provides for condition monitoring of the BWR penetrations.

The BWR water chemistry is controlled per the EPRI guidelines of BWRVIP-190 Revision 1 (TR-3002002623), BWR Water Chemistry Guidelines - Mandatory, Needed, and Good Practice Guidance (April 2014). BWRVIP-190 Revision 1 supersedes previous revisions of the guidelines for water chemistry in BWRs, including BWRVIP-29 (TR-103515), BWR Water Chemistry Guidelines - 1993 Revision, Normal and Hydrogen Water Chemistry (February 1993).

Program activities incorporate the inspection and evaluation guidelines of BWRVIP-49, BWR Vessel and Internals Project, Instrument Penetration Inspection and Flaw Evaluation Guidelines, for instrument penetrations and BWRVIP-27, BWR Vessel and Internals Project, BWR Standby Liquid Control System/Core Plate DP Inspection and Flaw Evaluation Guidelines, for the Standby Liquid Control System.

The program is updated periodically as required by 10 CFR 50.55a and the BWRVIP.

K2.1.10 BWR Stress Corrosion Cracking Program Description The Monticello Nuclear Generating Plant BWR Stress Corrosion Cracking Program is an existing program and is part of the MNGP ASME Section XI In-Service Inspection Program. ASME Section XI is being implemented with ultrasonic (UT) volumetric, surface, and visual inspections. NUREG-0313, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, and Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, and its Supplement 1 are part of the MNGP BWR Stress Corrosion Cracking Program. All IGSCC susceptible materials have been replaced or protected with a cladding of resistant weld material. Therefore, all piping welds are now classified as IGSCC Category A in accordance with NUREG-0313 and GL 88-01. As part of the MNGP recirculation piping replacement effort, austenitic stainless steel portions of piping systems 4" in nominal diameter or larger operating at temperatures above 200°F of the reactor coolant pressure boundary were replaced in accordance with the requirements of NUREG-0313.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 11 of 84 In addition, a Hydrogen Water Chemistry System was placed in operation, which reduces the oxidizing environment by introducing excess hydrogen to the reactor coolant system that combines with the free oxygen produced by radiolysis.

MNGP has also implemented use of Online Noble Metal (OLN) Chemistry to supplement water chemistry and aid in protection against SCC/IGSCC.

K2.1.11 BWR Vessel ID Attachment Welds Program Description The MNGP BWR Vessel ID Attachment Welds Program is part of the MNGP ASME Section XI In-Service Inspection Aging Management Program. The BWR Vessel ID Attachment Weld Program is in accordance with the approved ASME Section XI Edition / Addenda. The program provides for condition monitoring of the BWR vessel ID attachment welds. The program includes inspection and flaw evaluation in accordance with BWRVIP-48, Vessel ID Attachment Weld and Inspection and Flaw Guidelines (EPRI TR-108724).

The BWR water chemistry is controlled per the EPRI guidelines of BWRVIP-190 Revision 1 (TR-3002002623), BWR Water Chemistry Guidelines - Mandatory, Needed, and Good Practice Guidance (April 2014). BWRVIP-190 Revision 1 supersedes previous revisions of the guidelines for water chemistry in BWRs, including BWRVIP-29 (TR-103515), BWR Water Chemistry Guidelines - 1993 Revision, Normal and Hydrogen Water Chemistry (February 1993). This is considered an exception to the NUREG-1801 Program Description.

The Program is updated periodically as required by 10 CFR 50.55a. In addition the Program is supplemented by implementing the guidelines of Boiling Water Reactor Vessel and Internals Project (BWRVIP) documents.

K2.1.12 BWR Vessel Internals Program Description The MNGP BWR Vessel Internals Program is part of the MNGP ASME Section XI In-Service Inspection Program. The BWR Vessel Internals Program is in accordance with the approved ASME Section XI Edition / Addenda and approved alternatives included in the ISI Program. The program provides for condition monitoring of the BWR vessel internals for crack initiation and growth.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 12 of 84 MNGP activities include the in-vessel examination procedures and the plant water chemistry procedures. The in-vessel examination procedures implement the recommendations of the BWRVIP guidelines, as well as the requirements of Section XI of the ASME Boiler and Pressure Vessel Code and approved alternatives included in the ISI Program. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below system-specific limits based on the EPRI guidelines of BWRVIP-190 Revision 1 (TR-3002002623), BWR Water Chemistry Guidelines - Mandatory, Needed, and Good Practice Guidance (April 2014). BWRVIP-190 Revision 1 supersedes previous revisions of the guidelines for water chemistry in BWRs, including BWRVIP-29 (TR-103515), BWR Water Chemistry Guidelines - 1993 Revision, Normal and Hydrogen Water Chemistry (February 1993).

The Program is updated periodically as required by 10 CFR 50.55a and the BWRVIP Program.

Commitments Prior to the period of extended operation, the repair/replacement guidelines in BWRVIP-16, 19, 44, 45, 50, 51, 52, 57, and 58 were added, as applicable, to the MNGP BWR Vessel Internals Program.

During the period of extended operation, NSPM will perform top guide grid inspections using the EVT-1 method of examination, for the high fluence locations (grid beam and beam-to-beam crevice slot locations with fluence exceeding 5.0 x 1020 n/cm2). Ten percent (10%) of the total population will be inspected within 12 years with a minimum of 5% inspected within the first 6 years.

NSPM has inspected the in-core monitoring dry tubes on an every-other refueling outage periodicity and will continue to perform this inspection during the period of extended operation, per the guidance provided in GE SIL- 409.

The NSPM is an active member of the BWRVIP and will continue to follow applicable inspection guidelines and recommendations, which have been reviewed and approved by the executive committee of the BWRVIP, throughout the period of extended operation for MNGP.

NSPM will follow the guidance provided in BWRVIP-139, Steam Dryer Inspection and Flaw Evaluation Guidelines (April 2005), for the MNGP steam dryer inspections.

In the event a new steam dryer is installed, NSPM will reevaluate the inspection requirements. In 2011, a new steam dryer was installed at MNGP that is not addressed in BWRVIP-139. NSPM developed inspection requirements based on the manufacturer recommendations. Pre-EPU inspections were performed in RFO-26 (2013). Post-EPU inspections began in RFO-28 (2017).

NSPM added inspection requirements for the P1, P2, and P3 welds, at MNGP, in accordance with guidance provided in BWRVIP-18, or subsequent revisions.

K2.1.13 Closed-Cycle Cooling Water System Program Description The MNGP Closed-Cycle Cooling Water System Program includes: (1) preventive measures to minimize corrosion, and (2) periodic system and component performance testing and inspection to monitor the effects of corrosion and confirm intended

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 13 of 84 functions are met. Preventive measures include the monitoring and control of corrosion inhibitors and other chemical parameters, such as pH, in accordance with the guidelines of Electric Power Research Institute (EPRI) 3002000590, Closed Cooling Water Chemistry Guideline Revision 2, vendor recommendations, and plant operating experience. EPRI 3002000590 is the current revision (Revision 2) of EPRI-107396 and EPRI TR-1007820. As only minor changes were made to the MNGP Closed-Cycle Cooling Water System Program to implement EPRI 3002000590, the program is also still in accordance with the EPRI Revision 0 guidelines identified in NUREG-1801, Chapter XI Program M21, i.e., EPRI TR-107396, Closed Cooling Water Chemistry Guidelines. Periodic inspection and testing to confirm function and monitor corrosion is also performed in accordance with EPRI 3002000590, vendor recommendations, and industry and plant operating experience. A review of plant operating experience demonstrates these measures ensure closed-cycle cooling water (CCCW) systems are performing their intended functions.

The MNGP has four systems in License Renewal Scope that meet the definition for consideration as closed-cycle cooling water systems and portions of three additional systems (heat exchangers or coolers) that are serviced directly by these cooling water systems. These systems and portions of systems are not subject to significant sources of contamination, in which water chemistry is controlled and in which heat is not directly rejected to a heat sink. The adequacy of chemistry control is confirmed on a routine basis by sampling and monitoring to within established limits and by equipment performance monitoring to identify aging effects.

Corrosion inhibitor concentrations are maintained within limits based on a combination of EPRI 3002000590 guidelines, vendor recommendations, and plant experience.

System and component performance test results are evaluated in accordance with the guidelines of EPRI 3002000590 and used as a basis for evaluating the effectiveness of actions to mitigate cracking, corrosion, and heat exchanger fouling. Acceptance criteria and tolerances are also based on system design parameters and functions.

For chemical parameters monitored, many are based on ranges identical to or more restrictive than noted in EPRI 3002000590, EPRI TR-1007820, and EPRI TR-107396.

Others are based on vendor recommendations and plant experience.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 14 of 84 Frequency of performance and functional tests are consistent with EPRI 3002000590 and are based on plant operating experience, trends and equipment performance.

System and component operability tests are typically performed on a more frequent basis than once per cycle whereas more intrusive inspections (disassembly, eddy current testing, etc.) are performed less frequently but at sufficient intervals to detect the impact of aging effects on component function.

Commitments Prior to the period of extended operation, a one time inspection has been performed to monitor the effects of corrosion on select portions of closed-cycle cooling water systems that perform a pressure-integrity intended function.

K2.1.14 Compressed Air Monitoring Program Description The MNGP Compressed Air Monitoring Program consists of inspection, monitoring, and testing of the Instrument and Service Air System to provide reasonable assurance that it will perform their intended function for the duration of extended operation.

Commitments Prior to the period of extended operation, the Compressed Air Monitoring Program procedures have been revised to include corrective action requirements if the acceptance limits for water vapor, oil content, or particulate are not met. Also, the acceptance criteria for oil content testing will be clarified and the basis for the acceptance limits for the water vapor, oil content, and particulate tests will be provided.

Prior to the period of extended operation, the Compressed Air Monitoring Program has been revised to include inspection of air distribution piping based on the recommendations of EPRI TR-108147.

K2.1.15 Electrical Cables & Connections Not Subject To 10 CFR 50.49 Environmental Qualification Requirements Program Program Description The MNGP Electrical Cables & Connections Not Subject To 10 CFR 50.49 Environmental Qualification Requirements Program is a new program that manages the aging of conductor insulation material on cables, connectors, and other electrical insulation materials that are installed in an adverse localized environment caused by heat, radiation, or moisture. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for the component. An adverse variation in environment is significant if it could appreciably increase the rate of aging of a component or have an immediate adverse effect on operability.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 15 of 84 In most areas of the plant, the actual ambient environments (e.g., temperature, radiation, or moisture) are less severe than the plant design environment. However, in a limited number of localized areas, the actual environments may be more severe than the plant design environment for those areas. Cable and connection insulation materials may degrade more rapidly than expected in these adverse localized environments.

As stated in NUREG/CR-5643, The major concern with cables is the performance of aged cable when it is exposed to accident conditions. The statement of considerations for the final license renewal rule (60 Fed. Reg. 22477) states, The major concern is that failures of deteriorated cable systems (cables, connections, and penetrations) might be induced during accident conditions. Since they are not subject to the environmental qualification requirements of 10 CFR 50.49, the electrical cables and connections covered by this aging management program are either not exposed to harsh accident conditions or are not required to remain functional during or following an accident to which they are exposed.

The scope of this program includes accessible non-EQ electrical cables and connections, including control and instrumentation circuits, within the scope of license renewal.

The program provides reasonable assurance that the intended functions of electrical cables and connections within scope of license renewal that are not subject to the environmental qualification requirements of 10 CFR 50.49 and are exposed to adverse localized environments caused by heat, radiation, or moisture are maintained consistent with the current licensing basis through the period of extended operation.

This program considers the technical information and guidance provided in NUREG/CR-5643, IEEE Std. P1205-2000, SAND96-0344, and EPRI TR-109619.

The program addresses cables and connections whose configuration is such that most cables and connections installed in adverse localized environments are accessible. This program is a sampling program in which selected cables and connections from accessible areas are inspected and represent, with reasonable assurance, all cables and connections in the adverse localized environments. If an unacceptable condition or situation is identified for a cable or connection in the inspection sample, a determination is made as to whether the same condition or situation is applicable to other accessible or inaccessible cables or connections.

Commitments Prior to the period of extended operation, the MNGP Electrical Cables & Connections Not Subject To 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented as a new program consistent with the recommendations of NUREG-1801 Chapter XI Program XI.E1. The program will manage the aging of conductor insulation material on cables, connectors, and other electrical insulation materials that are installed in an adverse localized environment caused by heat, radiation, or moisture.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 16 of 84 K2.1.16 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrument Circuits Program Description This program applies to non-EQ electrical cables used in radiation monitoring and nuclear instrumentation circuits with sensitive, low-level signals that are within scope of license renewal and are installed in adverse localized environments caused by heat, radiation and moisture in the presence of oxygen.

In most areas within a nuclear power plant, the actual ambient environments (e.g.,

temperature, radiation, or moisture) are less severe than the plant design environment. However, in a limited number of localized areas, the actual environments may be more severe than the plant design environment for those areas. Conductor insulation materials used in electrical cables may degrade more rapidly than expected in these adverse localized environments. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for the cable. An adverse variation in environment is significant if it could appreciably increase the rate of aging of a component or have an immediate adverse effect on operability.

Exposure of electrical cables to adverse localized environments caused by heat or radiation can result in reduced insulation resistance (IR). Reduced IR causes an increase in leakage currents between conductors and from individual conductors to ground. A reduction in IR is a concern for circuits with sensitive, low-level signals such as radiation monitoring and nuclear instrumentation since it may contribute to inaccuracies in the instrument loop.

The purpose of the aging management program is to provide reasonable assurance that the intended functions of electrical cables that are not subject to the environmental qualification requirements of 10 CFR 50.49 and are used in circuits with sensitive, low-level signals exposed to adverse localized environments caused by heat, radiation or moisture will be maintained consistent with the current licensing basis through the period of extended operation. This program considers the technical information and guidance provided in NUREG/CR-5643, IEEE Std. P1205, SAND96-0344, and EPRI TR-109619.

In this aging management program, routine calibration tests performed as part of the plant surveillance test program are used to identify the potential existence of aging degradation. When an instrumentation loop is found to be out of calibration during routine surveillance testing, troubleshooting is performed on the loop, including the instrumentation cable.

In cases where a calibration or surveillance program does not include the cabling system in the testing circuit, or as an alternative to the review of calibration results described above, NSPM will perform cable system testing. A proven cable system test for detecting deterioration of the insulation system (such as insulation resistance tests, time domain reflectometry test, or other testing judged to be effective in determining cable insulation condition) will be performed.

As stated in NUREG/CR-5643, The major concern with cables is the performance of aged cable when it is exposed to accident conditions. The statement of considerations for the final license renewal rule (60 Fed. Reg. 22477) states, The

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 17 of 84 major concern is that failures of deteriorated cable systems (cables, connections, and penetrations) might be induced during accident conditions. Since they are not subject to the environmental qualification requirements of 10 CFR 50.49, the electrical cables covered by this aging management program are either not exposed to harsh accident conditions or are not required to remain functional during or following an accident to which they are exposed.

Commitments Prior to the period of extended operation, the Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Program will be implemented as a new program. With exceptions, it will be consistent with the recommendations of NUREG-1801 Chapter Xl, Program XI.E2.

K2.1.17 Fire Protection Program Description For license renewal purposes the MNGP Fire Protection Program includes a fire barrier inspection program, a diesel-driven fire pump inspection program, and a halon fire suppression system inspection.

The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of associated fire rated doors to ensure that their operability is maintained.

The diesel-driven fire pump inspection program requires that the pump be periodically tested and the diesel engine inspected to ensure that the fuel supply line can perform the intended function.

The halon fire suppression system inspection included periodic inspection and testing of the cable spreading room halon fire suppression system.

Commitments Prior to the period of extended operation:

1) The MNGP Fire Protection Program was revised to include a visual inspection of the halon fire suppression system to detect any signs of degradation, such as corrosion and mechanical damage. This visual inspection will provide aging management for external surfaces of the halon fire suppression system; and

2) Fire Protection Program plan document was revised to include qualification criteria for individuals performing visual inspections of penetration seals, fire barriers, and fire doors. The qualification criteria are based on physical capability (i.e., eye exam), education/training, and experience in the Fire Protection Program requirements.

K2.1.18 Fire Water System Program Description The Fire Water System aging management program relies on testing of water based fire protection system piping and components in accordance with applicable NFPA recommendations. In addition, this program will be modified to include (1) portions of the fire protection sprinkler system that are subjected to full flow tests prior to the period of extended operation and (2) portions of the fire protection system exposed to water that are internally visually inspected. To ensure that the aging mechanisms of

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 18 of 84 corrosion, and biofouling/fouling are properly being managed in the fire water system, periodic full flow flush test and system performance test are conducted. The system is also normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions initiated.

Commitments Prior to the period of extended operation, the Fire Water System Program:

1) Implementing procedures have been revised to include the extrapolation of inspection results to below grade fire water piping with similar conditions that exist within the above grade fire water piping, and

2) The MNGP Fire Water System Program sprinkler heads will be inspected and tested per NFPA requirements. Per the NFPA code, the sprinkler heads will be tested or replaced when the sprinklers have been in service for 50 years.

Testing procedures shall be repeated at 10-year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner. If the sprinkler heads are replaced testing is not required.

3) Has verified procedures to be used for aging management activities of the Fire Water System apply testing in accordance with applicable NFPA codes and standards. Revise the relevant procedures as appropriate.

K2.1.19 Flow-Accelerated Corrosion Program Description The Flow-Accelerated Corrosion Program manages aging effects (loss of material) due to flow-accelerated corrosion (FAC) on the internal surfaces of carbon or low alloy steel piping, elbows, reducers, expanders, and valve bodies which contain high energy fluids (both single phase and two phase). The program implements the EPRI guidelines in NSAC-202L-R3. This program also requires the use of CHECWORKS as a predictive tool. Included in the program are (a) an analysis to determine FAC susceptible locations; (b) performance of limited baseline inspections; (c) follow-up inspections to confirm the predictions; and (d) repairing or replacing components, as necessary.

The MNGP Flow-Accelerated Corrosion Program includes the response made to GL 89-08, Erosion/Corrosion Induced Pipe Wall Thinning.

Commitments The Fleet Procedure for the Flow Accelerated Corrosion Inspection Program will be revised to include the accepted 87.5% of nominal pipe wall thickness for non-safety related piping as a trigger point for engineering analysis.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 19 of 84 As a result of reviews for newly identified SSCs in scope of license renewal aging management in accordance with 10 CFR 54.37(b), this program's scope has been 604000000815 expanded to include components within the boundary of the Alternate Source Term (AST) and the following component group(s) have had "Plateout and holdup of radioactive materials" added as an 'intended function':

• Piping and Fittings

• Valve Bodies K2.1.20 Fuel Oil Chemistry Program Description The Monticello Nuclear Generating Plant (MNGP) Fuel Oil Chemistry Program is an existing program using existing diesel fuel oil system procedures that encompass the NUREG-1801 program recommendations. The Fuel Oil Chemistry Program mitigates and manages aging effects on the internal surfaces of diesel fuel oil storage tanks and associated components in systems that contain diesel fuel oil. The program includes (a) surveillance and monitoring procedures for maintaining diesel fuel oil quality by controlling contaminants in accordance with applicable ASTM Standards; (b) periodic draining of water from diesel fuel oil tanks, if water is present; (c) periodic or conditional visual inspection of internal surfaces or wall thickness measurements (e.g.,

by UT) from external surfaces of diesel fuel oil tanks; and (d) one-time inspections of a representative sample of components in systems that contain diesel fuel oil.

Commitments Prior to the period of extended operation:

1) The MNGP procedures related to the Diesel Fuel Oil System have been revised to include requirements to check for general, pitting, crevice, galvanic, microbiological influenced corrosion (MIC), and cracking.

2) The MNGP Fuel Oil Chemistry Program procedures have been revised to require tank draining, cleaning, and inspection if deemed necessary based on the trends indicated by the results of the diesel fuel oil analysis, or as recommended by the system engineer based on equipment operating experience.

3) Develop or revise existing procedures in the MNGP Fuel Oil Chemistry Program to require periodic tank inspections of the diesel fuel oil tanks.

K2.1.21 Inaccessible Medium Voltage (2kV to 34.5kV) Cables Not Subject to 10 CFR 50.49 EQ Requirements Program Description The purpose of this aging management program is to demonstrate that inaccessible, non-EQ medium-voltage cables susceptible to aging effects caused by moisture and voltage stress will be adequately managed so that there is reasonable assurance that the cables will perform their intended function in accordance with the current licensing basis during the period of extended operation. The intended function of insulated cables and connections is to provide electrical connections to specified sections of an electrical circuit to deliver voltage, current or signals.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 20 of 84 Most electrical cables at MNGP are located in dry environments. However, some cables may be exposed to condensation and wetting in inaccessible locations, such as conduits, cable trenches, cable troughs, duct banks, underground vaults or direct buried installations. When an energized medium-voltage cable is exposed to wet conditions for which it is not designed, water treeing or a decrease in the dielectric strength of the conductor insulation can occur. This can potentially lead to electrical failure.

In this aging management program, periodic actions are taken to prevent cables from being exposed to significant moisture, such as inspecting for water collection in cable manholes and conduit, and draining water, as needed. In-scope, medium-voltage cables exposed to significant moisture and significant voltage are tested to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined prior to the initial test, and is to be a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, polarization index, or other testing that is state-of-the-art at the time the test is performed.

Commitments Prior to the period of extended operation, the MNGP Inaccessible Medium Voltage (2 kV to 34.5 kV) Cables Not Subject to 10 CFR 50.49 EQ Requirements Program will be implemented as a new program consistent with the recommendations of NUREG-1801 Chapter XI Program XI.E3.

K2.1.22 Inspection of Overhead Heavy Load & Light Load (Related to Refueling) Handling Systems Program Description The Inspection Of Overhead Heavy Load & Light Load (Related To Refueling)

Handling Systems program, which is implemented through plant procedures and preventive maintenance, manages loss of material of structural components for heavy load and fuel handling components within the scope of license renewal. The Inspection Of Overhead Heavy Load & Light Load (Related To Refueling) Handling Systems program provides for visual and NDE inspections of in-scope load handling components. Functional tests are also performed to assure their integrity. The cranes also comply with the maintenance rule requirements provided in 10 CFR 50.65.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 21 of 84 Commitments Prior to the period of extended operation, the Inspection of Overhead Heavy Load &

Light Load (Related to Refueling) Handling Systems Program has been enhanced to specify a five-year inspection frequency for the fuel preparation machines.

K2.1.23 One-Time Inspection Program Description The MNGP One-Time Inspection Program is a new program that is being developed consistent with NUREG-1801 Chapter XI Program M32, One-Time Inspection. Any exceptions or enhancements to NUREG-1801 will be described in the relevant element descriptions. This program includes measures to verify the effectiveness of the following aging management programs:

• Plant Chemistry Program

• Fuel Oil Chemistry Program This program also confirms the absence of age degradation in selected components (e.g., flow restrictors, venturis, and small bore piping) within License Renewal scope.

The MNGP One-Time Inspection Program addresses concerns and confirmation for the potential long incubation period for certain aging effects on structures and components. There are cases where either (a) an aging effect is not expected to occur but there is insufficient data to completely rule it out, or (b) an aging effect is expected to progress very slowly.

The activities of the One-Time Inspection Program include (a) determination of the sample size based on an assessment of materials of fabrication, environment, plausible aging effects, and operating experience; (b) identification of the inspection locations in the system or component based on the aging effect; (c) determination of the examination technique, including acceptance criteria that would be effective in managing the aging effect for which the component is examined; and (d) evaluation of the need for follow-up examinations to monitor the progression of any identified aging degradation.

The program will manage the aging effects due to corrosion, cracking, erosion, fouling, fretting, or thermal exposure. The program will also verify the absence of reduction of neutron absorption capacity of boral in the spent fuel pool.

Commitments Prior to the period of extended operation, the MNGP One-Time Inspection Program will be implemented as a new program consistent with the recommendations of NUREG-1801 Chapter Xl Program XI.M32, "One-Time Inspection." This program will include measures to verify the effectiveness of the following aging management programs: Plant Chemistry Program and Fuel Oil Chemistry Program. This program will also confirm the absence of age related degradation in selected components (e.g.,

flow restrictors, venturis) within License Renewal scope.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 22 of 84 K2.1.24 Open-Cycle Cooling Water System Program Description The MNGP Open-Cycle Cooling Water System Program relies on the implementation of the recommendations of NRC Generic Letter (GL) 89-13 to ensure that the effects of aging on the raw water service water systems will be managed for the period of extended operation. This program manages the aging effects of metallic components in water systems (e.g., piping and heat exchangers) exposed to raw, untreated (e.g.,

service) water. These aging effects are due to corrosion, erosion, and biofouling in systems, structures and components serviced by the OCCW system. The program includes (a) surveillance and control of biofouling; (b) tests to verify heat transfer; and (c) routine inspection and maintenance.

The MNGP Open-Cycle Cooling Water System Program complies with MNGP's response to NRC GL 89-13. Resultant commitments made to comply with GL 89-13 have been incorporated into plant procedures and programs.

K2.1.25 Plant Chemistry Program Program Description The MNGP Plant Chemistry Program mitigates the aging effects on component surfaces that are exposed to water as the process fluid; chemistry programs are used to control water chemistry for impurities (e.g., chloride and sulfate) that accelerate corrosion or crack initiation and growth and that cause heat transfer degradation due to fouling in select heat exchangers. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below system-specific limits based on BWRVIP-190 Revision 1 (TR-3002002623), BWR Water Chemistry Guidelines - Mandatory, Needed, and Good Practice Guidance (April 2014). BWRVIP-190 Revision 1 supersedes previous revisions of the guidelines for water chemistry in BWRs, including BWRVIP-29 (TR-103515), BWR Water Chemistry Guidelines - 1993 Revision, Normal and Hydrogen Water Chemistry (February 1993).

For low-flow or stagnant portions of a system, a one-time inspection of selected components at susceptible locations provides verification of the effectiveness of the Plant Chemistry Program.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 23 of 84 As a result of reviews for newly identified SSCs in scope of license renewal aging management in accordance with 10 CFR 54.37(b), this program's scope has been expanded to include components within the boundary of the Alternate Source Term (AST) and the following component group(s) have had "Plateout and holdup of radioactive materials" added as an 'intended function':

604000000815

• Heat Exchangers

• Piping and Fittings

• Pump Casings

• Tanks

• Valve Bodies K2.1.26 Primary Containment In-Service Inspection Program Program Description The MNGP Primary Containment In-Service Inspection Program requires visual examinations of the accessible surfaces (base metal and welds) of the drywell, torus, vent lines, internal vent system, penetration assemblies and associated integral attachments. The program also requires examination of pressure retaining bolting and the drywell interior slab moisture barrier.

The program conforms to the applicable requirements of 10 CFR 50.55a and the ASME Boiler and Pressure Vessel Code, Subsection IWE.

General visual examinations that assess overall structural condition are performed once during each period as modified by 10 CFR50.55a.

Surface and / or volumetric examination augments visual examination as required to define the extent of observed conditions or to identify deterioration at inaccessible locations.

Examinations are scheduled and performed as required to evaluate bolting when in place or disassembled, and the condition of the normally submerged torus surface.

The program is updated periodically as required by 10 CFR50.55a.

K2.1.27 Protective Coating Monitoring & Maintenance Program Program Description The MNGP Protective Coating Monitoring and Maintenance Program applies to Service Level 1 protective coatings inside containment to address the concerns of NRC GL 98-04, Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Cooling Accident because of Construction and Protective Coating Deficiencies and Foreign Material in Containment. The Protective Coating Monitoring and Maintenance Program prevents the degradation of coatings that could lead to the clogging of ECCS suppression pool suction strainers. MNGP does not credit the Protective Coating Monitoring and Maintenance Program for the prevention of corrosion of carbon steel components.

As outlined in MNGP's response to GL 98-04, the Protective Coating Monitoring and Maintenance Program is a comparable program for monitoring and maintaining

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 24 of 84 protective coatings inside the primary containment and subject to the requirements of ANSI N101.4-1972 as modified by Regulatory Guide 1.54, June 1973.

Commitments Prior to the period of extended operation, the MNGP Protective Coating Maintenance and Monitoring Program:

1) Procedures have been updated to include Inspection of all accessible painted surfaces inside containment;

2) Procedures have been revised to include a pre-inspection review of the previous two inspection reports so that trends can be identified;

3) Implementation procedures have been revised to include provisions for analysis of suspected reasons for coating failure; and

4) The MNGP Protective Coating Maintenance and Monitoring Program has been revised to include the requirement that personnel performing the inspection shall meet the requirements of ANSI N45.2.6.

K2.1.28 Reactor Head Closure Studs Program Description The MNGP Reactor Head Closure Studs Program is part of the MNGP ASME Section XI In-Service Inspection Program. The Reactor Head Closure Studs Program is in accordance with the approved ASME Section XI Edition / Addenda and provides for condition monitoring of the reactor head closure stud bolting. Replacement reactor head studs available for use at Monticello include preventive measures described in RG 1.65, Material and Inspection for Reactor Vessel Closure Studs. The Program is updated periodically as required by 10 CFR 50.55.a.

K2.1.29 Reactor Vessel Surveillance Program Description The MNGP Reactor Vessel Surveillance Program is part of the Boiling Water Reactors Vessel Internals Project (BWRVIP) Integrated Surveillance Program (ISP) that uses data from BWR member surveillance programs to select the best representative material to monitor radiation embrittlement for a particular plant. The BWRVIP ISP monitors capsule test results from various member plants. This is consistent with the methodology allowed by NUREG-1801.

The MNGP Reactor Vessel Surveillance Program is required by 10 CFR 50, Appendix H. The scope of the Reactor Vessel Surveillance Program is described by the BWRVIP ISP guidance. The ISP capsule removal schedule is included in BWRVIP-86 and its technical basis is described in BWRVIP-78.

The NRC in a Safety Evaluation (SE) to the BWRVIP, dated February 1, 2002, approved the ISP. This Safety Evaluation concluded that the ISP, if implemented in accordance with the conditions in the SE, is an acceptable alternative to all existing BWR plant-specific RPV surveillance programs for the purpose of maintaining compliance with the requirements of Appendix H to 10 CFR Part 50 through the end of current facility 40 year operating licenses.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 25 of 84 Commitments MNGP intends to use the Integrated Surveillance Program for MNGP during the period of extended operation by implementing the requirements of BWRVIP-86 R1-A, and subsequent NRC Approved Revisions.

Prior to and during the period of extended operation, MNGP will retain the capsules removed from the MNGP reactor vessel as part of the Reactor Vessel Surveillance Program.

K2.1.30 Selective Leaching of Materials Program Description The MNGP Selective Leaching of Materials Program was a new program, developed and implemented before the start of the period of extended operation. The program included one-time visual inspection and hardness measurement of selected components that were susceptible to selective leaching. The results of these inspections determined that selective leaching is occurring in cast iron components.

Based on these results, the program was developed into an ongoing opportunistic inspection program of cast iron components.

Opportunistic inspections are performed when scheduled work is performed on susceptible components. The inspections include visual and hardness measurement of the susceptible component. In situations where hardness testing is not practical, a qualitative method by other NDE or metallurgical methods will be used to determine the presence and extent of selective leaching.

Commitments Prior to the period of extended operation, the MNGP Selective Leaching of Materials Program was implemented as a new program consistent, with exceptions, to the recommendations of NUREG-1801 Chapter Xl Program XI.M33, "Selective Leaching of Materials." The program was developed and implemented before the start of the period of extended operation. The program included a one-time visual inspection and hardness measurement of selected components that are susceptible to selective leaching. In situations where hardness testing was not practical, a qualitative method by other NDE or metallurgical methods was used to determine the presence and extent of selective leaching. The program was to determine if selective leaching is occurring for selected components.

K2.1.31 Structures Monitoring Program Program Description The MNGP Structures Monitoring Program provides for aging management of structures and structural components within the scope of license renewal and implements the NUREG-1801, XI.S6, Structures Monitoring Program. The Structures Monitoring Program is based on the guidance provided in RG 1.160 and NUMARC 93-

01. The Structures Monitoring Program is implemented as part of the structures monitoring done under the MNGP Maintenance Rule Program and with additional inspections of the intake structure and diesel fuel oil transfer house.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 26 of 84 The Structures Monitoring Program also implements the NUREG-1801, XI.S5, Masonry Wall Program. Masonry block wall inspections are performed as part of the maintenance rule inspections and are based on IEB 80-11 with administrative controls per IN 87-67.

As permitted by NUREG-1801, XI.S7, RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants, the inspection of water control structures is included in the Structures Monitoring Program. The only water control structure in scope for license renewal is the intake structure. Maintenance rule inspections are performed on the portions of the intake structure above the water line.

The Structures Monitoring Program includes separate inspections of the underwater portions of the intake structure.

In addition, special settlement checks of the diesel fuel oil transfer house are performed outside the Maintenance Rule inspections.

The Structures Monitoring Program does not rely upon protective coatings to manage the effects of aging.

Commitments Prior to the period of extended operation, the MNGP Structures Monitoring Program:

1) Was expanded, as necessary, to include inspections of structures and structural elements in scope for License Renewal that are not inspected as part of another aging management program.

2) Implementing procedures have been enhanced to ensure that structural inspections are performed on submerged portions of the intake structure from the service water bays to the wing walls;

3) Implementing procedures have been revised to include the monitoring/inspection parameters for structural components within the scope of License Renewal;

4) Was enhanced to include a requirement to sample ground water for pH, chloride concentration and sulfate concentration;

5) Was enhanced to include concrete evaluations of inaccessible areas if degradation of accessible areas is detected; and

6) Implementing procedures have been enhanced to include acceptance criteria for structural inspections of submerged portions of the Intake Structure.

604000000841 The Structures Monitoring Program also implements portions of the NUREG-2191, GALL-SLR Report,Section XI.M29. The concrete foundation and sealant at the base of the CST tanks are managed by the MNGP Structures Monitoring AMP.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 27 of 84 K2.1.32 System Condition Monitoring Program Program Description The System Condition Monitoring Program is an existing plant-specific program that is based on system engineer monitoring. Although many monitoring activities are being performed at MNGP, this AMP brings aging management into the scope of the monitoring activities. Other groups augment this program by identifying and reporting adverse material conditions via the corrective action process or work control process.

This monitoring consists of system-level performance monitoring, inspections and walkdowns, health and status reporting, and preventive maintenance. This program has been enhanced to include specific activities and criteria for managing age related degradation for SSCs within License Renewal scope. This program manages aging effects for normally accessible external surfaces of piping, tanks, hangers/supports, racks, panels, and other components and equipment within the scope of License Renewal. These aging effects are managed through visual inspection and monitoring of external surfaces for leakage and evidence of material degradation.

Commitments Prior to the period of extended operation, the implementing instructions and procedures for the MNGP System Condition Monitoring Program were revised to describe specific age degradation parameters to be monitored and inspected.

Acceptance criteria will also be included.

As a result of reviews for newly identified SSCs in scope of license renewal aging management in accordance with 10 CFR 54.37(b), this program's scope has been expanded to include components within the boundary of the Alternate Source Term (AST) and the following component group(s) have had "Plateout and holdup of radioactive materials" added as an 'intended function':

604000000815

• Heat Exchangers

• Piping and Fittings

• Pump Casings

• Tanks

• Valve Bodies

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 28 of 84 K2.1.33 Thermal Aging & Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)

Program Description The MNGP Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program monitors the aging effects of loss of fracture toughness on the intended function of the component by performing examinations on CASS reactor vessel internal components as part of the MNGP ASME Section XI In-Service Inspection Program. The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program is in accordance with ASME Section XI, Subsection IWB, Category B-N-1 and B-N-2 requirements and provides for condition monitoring of the CASS components. Additional enhanced visual inspections that incorporate the requirements of the BWRVIP are performed to detect the effects of loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of CASS reactor vessel internals.

The program is updated periodically as required by 10 CFR 50.55a.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 29 of 84 K2.1.34 Electrical Cable Connections Not subject to 10 CFR 50.49 Environmental Qualification Requirements Program Description Cable connections are used to connect cable conductors to other cables or electrical devices. Connections associated with cables within the scope of license renewal are part of this program. The most common types of connections used in nuclear power plants are splices (butt or bolted), crimp-type ring lugs, connectors, and terminal blocks. Most connections involve insulating material and metallic parts. This aging management program for electrical cable connections (metallic parts) account for the following aging stressors: thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, and oxidation.

GALL XI.E1, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, manages the aging of insulating material but not the metallic parts of the electrical connections. GALL XI.E1 is based on only a visual inspection of accessible cables and connections. Visual inspection is not sufficient to detect the aging effects from thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, and oxidation on the metallic parts of cable connections.

Circuits exposed to appreciable ohmic or ambient heating during operation may experience loosening related to repeated cycling of connected loads or of the ambient temperature environment. Different materials used in various cable system components can produce situations where stresses existing between these components change with repeated thermal cycling. For example, under loaded conditions, appreciable ohmic heating may raise the temperature of a compression termination and cable conductor well above the ambient temperature, thereby causing thermal expansion of both components. Different thermal expansion coefficients may alter mechanical stresses between the components so that the termination may tighten on the conductor. When the load or current is reduced, the affected components cool and contract. Repeated cycling in this fashion can produce loosening of the termination under ambient conditions, and may lead to high electrical resistance joints or eventual separation to compression-type terminations. Threaded connectors, splices, and terminal blocks may loosen if subjected to significant thermally induced stress and cycling.

Cable connections within the scope of license renewal are tested to provide an indication of the integrity of the cable connections. The specific type of test performed will be determined prior to the initial test, and will be a proven test for detecting loose connections, such as thermography, contact resistance testing, or other appropriate testing.

This program, as described, can be thought of as a sampling program. The following factors are considered for sampling: application (high, medium and low voltage),

circuit loading, and location (high temperature, high humidity, vibration, etc.). The technical basis for the sample selections is documented. If an unacceptable condition or situation is identified in the selected sample, a determination is made as to whether the same condition or situation is applicable to other connections not tested.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 30 of 84 SAND 96-0344, Aging Management Guidelines for Electrical Cable and Terminations, indicated loose terminations were identified by several plants. The major concern is that failures of a deteriorated cable system (cables, connections including fuse holders, and penetrations) might be induced during accident conditions.

Since the connections are not subject to the environmental qualification requirements of 10 CFR 50.49, an aging management program is required to manage the aging effects. This program will ensure that electrical cable connections will perform their intended function for the period of extended operation.

K2.1.35 Outdoor and Large Atmospheric Metallic Storage Tanks (OLAMST)

Program Description The MNGP Outdoor and Large Atmospheric Metallic Storage Tanks (OLAMST) AMP is a new aging management program that was created as a result of a review of the licensing basis for Station Black Out (SBO) in support of the Subsequent License Renewal (SLR) project. It was discovered that T-1A and T-1B, Condensate Storage Tanks (CSTs), should be considered for inclusion within scope of the current (first) license renewal (QIM 501000054498).

The MNGP Outdoor and Large Atmospheric Metallic Storage Tanks AMP implements NUREG-2191,Section XI.M29 and will manage the effects of aging on the external and internal surfaces of Condensate Storage Tanks (CSTs), T-1A and T-1B, so that component intended function(s) are maintained consistent with the current licensing basis (CLB) for the period of extended operation (PEO).

This AMP will manage loss of material by conducting periodic internal and external visual inspections. Inspections of caulking or sealant are supplemented with physical 604000000841 manipulation. Thickness measurements of tank bottoms are conducted to detect degradation (e.g., loss of material on the inaccessible external surface). The external surfaces of insulated tanks will be periodically inspected using a sampling of inspection points. Internally coated/lined surfaces are also managed for the loss of coating or lining integrity in this program. The concrete foundation and sealant at the base of the tanks are managed by the MNGP Structures Monitoring AMP. The tank internals exposed to treated water including tank walls and bottoms were inspected in 2017.

Inspections and tests will be performed by personnel qualified in accordance with site procedures and programs to perform the specified task. Inspections and tests will follow site procedures that include inspection parameters for items such as lighting, distance, offset, surface coverage, presence of protective coatings, and cleaning processes. Periodic external visual inspections will be performed each outage to confirm that insulation, insulation jacketing, sealant, and caulking are intact.

Inspections include locations where potential leakage past the insulation could be accumulating. The visual inspection of sealant and caulking are supplemented with physical manipulation to detect degradation. The visual inspections of the tank internals exposed to air and condensation environment, and the tank bottom thickness measurements will be performed each 10-year period

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 31 of 84 Where practical, identified degradation will be projected until the next scheduled inspection. Results will be evaluated against acceptance criteria to confirm that the timing of subsequent inspections will maintain the component intended function throughout the PEO based on the projected rate of degradation. Applicable operating experience or inspection results may also be used to justify performing the periodic inspections more frequently.

604000000841 Acceptance criteria for the tank to concrete interface inspection are to confirm there are no signs of aging as a result of water intrusion. Any degradation of paints, coatings, or evidence of corrosion will be reported, and further evaluation is required to determine if repair or replacement will be conducted. When the acceptance criteria are not met, the condition is entered into the corrective action program for evaluation and determination of corrective actions to ensure that the intended functions of the tanks will be maintained under all CLB design conditions. Results will be evaluated against acceptance criteria for the in-scope tanks to confirm that the timing of the subsequent inspections will maintain the component intended functions throughout the PEO based on the projected rate of degradation.

K3 Evaluation of Time-Limited Aging Analyses As part of a License Renewal Application, 10 CFR 54.21(c) requires that an evaluation of time-limited aging analyses (TLAAs) for the period of extended operation be provided. The following TLAAs have been identified and evaluated to meet this requirement.

Where used herein, the term rerate refers to the MNGP Power Rerate Program, which resulted in an increase in rated thermal power from 1670 MWt to 1775 MWt (approximately 6.3 percent). The increase in rated thermal power was implemented at MNGP in 1998. To demonstrate margin, most analyses performed for the power rerate conservatively used a power level of 1880 MWt. The continued use of this conservatism is described, where appropriate, in the following TLAA evaluations.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 32 of 84 K3.1 Neutron Embrittlement of the Reactor Pressure Vessel and Internals The materials of the reactor pressure vessel (RPV) and internals are subject to embrittlement due to high energy (E > 1 MeV) neutron exposure. Embrittlement means the material has lower toughness (i.e., will absorb less strain energy during a crack or rupture), thus allowing a crack to propagate more easily under thermal and/or pressure loading.

Toughness (indirectly measured in foot-pounds of absorbed energy in a Charpy impact test) is temperature-dependent in ferritic materials. An initial nil-ductility reference temperature (RTNDT), the temperature associated with the transition from ductile to brittle behavior, is determined for vessel materials through a combination of Charpy and drop weight testing. Toughness increases with temperature up to a maximum value called the upper-shelf energy (USE). Neutron embrittlement causes an increase in the RTNDT and a decrease in the USE of RPV steels. The increase or shift in the initial nil-ductility reference temperature (RTNDT) means higher temperatures are required for the material to continue to act in a ductile manner.

To reduce the potential for brittle fracture during RPV operation by accounting for the changes in material toughness as a function of neutron radiation exposure (fluence),

operating pressure-temperature (P-T) limit curves are included in the licensee-controlled Pressure Temperature Limits Report (PTLR) (Reference 28). The P-T curves account for the decrease in material toughness associated with a given fluence, which is used to predict the loss in toughness of the RPV materials. Based on the projected drop in toughness for a given fluence, the P-T curves are generated to provide a minimum temperature limit associated with the vessel pressure. The P-T curves are determined by the RTNDT and RTNDT values for the licensed operating period along with appropriate margins.

The RPV RTNDT and USE, calculated on the basis of neutron fluence, are part of the licensing basis and support safety determinations. Therefore, these calculations are Time-Limited Aging Analyses (TLAAs). The increases in RTNDT (RTNDT) affect the bases for relief from circumferential weld inspection and their associated supporting calculation of limiting axial weld conditional failure probability. As such, circumferential weld examination relief and axial weld failure probability are also TLAAs. Section K3.1 includes the following TLAA discussions related to the issue of neutron embrittlement:

• RPV Materials USE Reduction Due to Neutron Embrittlement

• Adjusted Reference Temperature (ART) for RPV Materials Due to Neutron Embrittlement

• Reflood Thermal Shock Analysis of the RPV

• Reflood Thermal Shock Analysis of the RPV Core Shroud

• RPV Thermal Limit Analysis: Operating P-T Limits

• RPV Circumferential Weld Examination Relief

• RPV Axial Weld Failure Probability

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 33 of 84 RPV Materials USE Reduction Due to Neutron Embrittlement Summary Description USE is the standard industry parameter used to indicate the maximum toughness of a material at high temperature. 10 CFR 50 Appendix G requires the predicted end-of-life Charpy impact test USE for RPV materials to be at least 50 ft/lb (absorbed energy), unless an approved analysis supports a lower value. Initial unirradiated test data are available for only one plate heat for the MNGP RPV to demonstrate a minimum 50 ft-lb USE by standard methods. End-of-life fracture energy was evaluated by using an equivalent margin analysis (EMA) methodology approved by the NRC for all other materials (Reference 1). This analysis confirmed that an adequate margin of safety against fracture, equivalent to 10 CFR 50 Appendix G requirements, does exist. The end-of-life USE calculations satisfy the criteria of 10 CFR 54.3(a) (Reference 2). As such, these calculations are a TLAA.

Analysis The MNGP RPV was designed for a 40-year life with an assumed neutron exposure of less than 1019 n/cm2 from energies exceeding 1 MeV. The current licensing basis calculations use realistic calculated fluences that are lower than this limiting value. The design basis value of 1019 n/cm2 bounds calculated fluences for the original 40-year term.

The tests performed on RPV materials under the Code of Record provided limited Charpy impact data. It was possible to develop original Charpy impact test USE values for only one plate material using the methods of 10 CFR 50 Appendix H and American Society For Testing and Materials (ASTM) E185 invoked by 10 CFR 50 Appendix G. Therefore, alternative methods approved by the NRC in NEDO-32205-A, have been used to demonstrate compliance with the 40-year 50 ft-lb USE requirement.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 34 of 84 Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

Fluence was calculated for the MNGP RPV for the extended 60-year (54 EFPY) licensed operating periods, using the methodology of NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation, which was approved by the NRC in a letter dated September 14, 2001 from S.A.

Richards (NRC) to J.F. Klapproth (GE) (Reference 3). The NRC found that, in general, this methodology adheres to the guidance in Regulatory Guide 1.190 for neutron flux evaluation. The fluence calculation was updated for operation at 2004 MWt using the NEDC-32983P-A methodology and accepted for use by the NRC under license amendment 176 as described in Reference 29. Peak fluence was calculated at the RPV inner surface (inner diameter), for purposes of evaluating USE. The value of neutron fluence was also calculated for the 1/4T location into the RPV wall measured radially from the inside diameter (ID), using Equation 3 from Paragraph 1.1 of Regulatory Guide (RG) 1.99, Revision 2. This 1/4T depth is recommended in the ASME Boiler and Pressure Vessel Code Section XI, Appendix G Subarticle G-2120 as the maximum postulated defect depth.

The End of License (EOL) USE was evaluated by an EMA using the 54 EFPY calculated fluence, and MNGP surveillance capsule results. As described in the Safety Evaluation Report (SER) to BWRVIP-74 (Reference 4), the percent reduction in Charpy USE for the limiting BWR/3-6 plates and BWR/2-6 welds are 23.5% and 39% respectively. Table K3.1-1 and Table K3.1-2 provide results of the EMA for limiting welds and plates on the RPV. As noted in Table K.3-1-1, plate C2220-2 (1-14) is not bounded by EMA but is shown to be acceptable by an alternate method. With a transverse unirradiated USE of 86.5 ft-lb (0.65 x 133 ft-lbs), a 33.5 percent reduction results in a 54 EFPY Charpy USE of 57.5 ft-lb, which exceeds the 50 ft-lb minimum identified in Appendix G to 10 CFR Part 50.

Because the projected USE exceeds the minimum recommended by Appendix G to 10 CFR Part 50, the 54 EFPY USE for this plate is acceptable (Ref. 22, Section 4.2.1.2). Refer to Ref. 22 for discussion of resolution of RAI 4.2-2 which discussed USE for vessel plates. The 54 EFPY USE values are managed in conjunction with surveillance capsule results as part of the BWRVIP Integrated Surveillance Program, BWRVIP-86-A (Reference 5) and BWRVIP-116 (Reference 15).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 35 of 84 Table K3.1-1 Equivalent Margin Analysis for MNGP Plate Material BWR/3-6 PLATE Surveillance Plate USE:

%Cu = 0.17 1st Capsule Fluence = 2.93 x 1017 n/cm2 2nd Capsule Fluence = 9.05 x 1017 n/cm2 1st Capsule Measured % Decrease = N/A (Charpy Curves) 2nd Capsule Measured % Decrease = N/A (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease = 11 (RG 1.99, Figure 2) 2nd Capsule RG 1.99 Predicted % Decrease = 14 (RG 1.99, Figure 2)

Limiting Beltline Plate USE:

%Cu = 0.17 54 EFPY 1/4T Fluence = 3.82 x 1018 n/cm2 RG 1.99 Predicted % Decrease = 21 (RG 1.99, Figure 2)

Adjusted % Decrease = 33.5 (Ref 22 Section 4.2.1.2) 33.5% > 23.5%, so vessel plate C2220-2 is not bounded by EMA

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 36 of 84 Table K3.1-2 Equivalent Margin Analysis for MNGP Weld Material BWR/3-6 WELD Surveillance Weld USE:

%Cu = 0.04 1st Capsule Fluence = 2.93 x 1017 n/cm2 2nd Capsule Fluence = 9.05 x 1017 n/cm2 1st Capsule Measured % Decrease = N/A (Charpy Curves) 2nd Capsule Measured % Decrease = N/A (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease = 8 (RG 1.99, Figure 2) 2nd Capsule RG 1.99 Predicted % Decrease = 11 (RG 1.99, Figure 2)

Limiting Beltline Weld USE:

%Cu = 0.10 54 EFPY 1/4T Fluence = 3.82 x 1018 n/cm2 RG 1.99 Predicted % Decrease = 19.5 (RG 1.99, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Position 2.2) 19.5 < 39%, so vessel welds are bounded by EMA.

For three components (C2220-2 plate, welds and N2 nozzles) the USE acceptance criteria is

> 50 ft-lb. The 54 EFPY USE for C2220-2 is 57.5 ft-lb. The 54 EFPY USE for the welds is 68 ft-lb. The 54 EFPY USE for the N2 nozzles is 52 ft-lb. Therefore, the 54 EFPY USE for affected vessel components is acceptable. (Ref 22, Table 4.2.1-1)

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 37 of 84 Adjusted Reference Temperature (ART) for RPV Materials Due to Neutron Embrittlement Summary Description The initial RTNDT, nil-ductility reference temperature, is the temperature at which a non-irradiated metal (ferritic steel) changes in fracture characteristics going from ductile to brittle behavior. RTNDT was evaluated according to the procedures in the ASME Code, Paragraph NB-2331. Neutron embrittlement raises the initial nil-ductility reference temperature. 10 CFR 50 Appendix G defines the fracture toughness requirements for the life of the vessel. The shift to the initial nil-ductility reference temperature (RTNDT) is evaluated as the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. This increase (RTNDT) means that higher temperatures are required for the material to continue to act in a ductile manner. The ART is defined as RTNDT + RTNDT + margin. The margin is defined in RG 1.99. The P-T curves are developed from the ART for the RPV materials. These are determined by the unirradiated RTNDT and by the RTNDT calculations for the licensed operating period. RG 1.99 defines the calculation methods for RTNDT, ART, and end-of-life USE.

The RTNDT and ART calculations meet the criteria of 10 CFR 54.3(a). As such, they are TLAAs.

Analysis The MNGP RPV was designed for a 40-year life with an assumed neutron exposure of less than 1019 n/cm2 from energies exceeding 1 MeV (Reference 6).

The current licensing basis calculations use realistic calculated fluences that are lower than this limiting value. The design basis value of 1019 n/cm2 bounds calculated fluences for the original 40-year term. The RTNDT values were determined using the embrittlement correlations defined in RG 1.99.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 38 of 84 Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

Fluence was calculated for the MNGP RPV for the extended 60-year (54 EFPY) licensed operating period, using the methodology of NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation, which was approved by the NRC in a letter dated September 14, 2001 from S.A.

Richards (NRC) to J.F. Klapproth (GE) (Reference 3). The NRC found that, in general, this methodology adheres to the guidance in Regulatory Guide 1.190 for neutron flux evaluation. The fluence calculation was updated for operation at 2004 MWt using the NEDC-32983P-A methodology and accepted for use by the NRC under license amendment 176 as described in Reference 29. Peak fluence was calculated at the vessel inner surface (inner diameter), for purposes of evaluating USE and ART. The value of neutron fluence was also calculated for the 1/4T location into the vessel wall measured radially from the inside diameter (ID), using Equation 3 from Paragraph 1.1 of RG 1.99. This 1/4T depth is recommended in the ASME Boiler and Pressure Vessel Code Section XI, Appendix G Sub-article G-2120 as the maximum postulated defect depth.

The 54 EFPY RTNDT for all beltline materials was calculated based on the embrittlement correlation found in RG 1.99. The peak fluence, RTNDT, and ART values for the 60-year (54 EFPY) license operating period are presented in Table K3.1-3. For evaluation of ART, additional margin was added to the peak fluence to account for potential future changes to operation during the current license.

The additional margin results in higher fluence for all vessel shell plates, the limiting beltline weld and the N2 nozzle. The margin was added within the bounds of NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (Reference 3). The use of this higher fluence is limited to the evaluation of the ART and RTNDT and is conservative with respect to the calculated fluence for the MNGP RPV at 54 EFPY (Reference 27). This table shows that the limiting ARTs allow P-T limits that will provide reasonable operational flexibility.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 39 of 84 The beltline region is defined as that portion of the RPV adjacent to the active fuel that attains a fluence 1.0 x 1017 n/cm2 during plant license. This extends the beltline 18" below and 168" above the bottom of active fuel (approximately 23" above the top of active fuel). As a result, the N2 Recirculation Inlet Nozzle falls within this extended beltline region, and is included in the calculation for ART in Table K3.1-3. In the absence of copper data for the N2 nozzle, this value is based upon heats of materials used for beltline nozzles at other plants (see Table K3.1-3). The nickel content has been determined as the average from all material test reports for the MNGP N2 nozzles. Additionally, the girth weld between Shell Rings 2 and 3 falls into the extended beltline region. The limiting weld values presented in Table K3.1-3 represent this girth weld in addition to the other vertical and girth welds in the beltline region. The ART values in Table K3.1-3 were used in the development of the current 54 EFPY P-T curves in the Monticello Nuclear Generation Plant PTLR. The current P-T curves in the PTLR were developed using methodology SIR-05-044-A, Pressure Temperature Limits Report Methodology for Boiling Water Reactors (Reference 25). The PTLR was approved by the NRC under Issuance of Amendment to Revise and Relocate Pressure Temperature Curves to a Pressure Temperature Limits Report (TAC NO. ME7930), dated February 27, 2013 (Reference 26).

The MNGP RTNDT and ART values are managed in conjunction with surveillance capsule results from the BWRVIP Integrated Surveillance Program, BWRVIP-86-A (Reference 5) and BWRVIP-116 (Reference 15).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 40 of 84 Table K3.1-3 60 Year Analysis Results for MNGP Chemistry Adjustments For 1/4t Lot Initial Chemistry Description Heat No. RTNDT Margin Terms ARTNDT Number RTNDT (°F) Cu (wt%) Ni (wt%) Factor (°F)

(°F) (°F) (°F) (°F)

Upper/Int Shell I-12 C2089-1 N/A 0.0 0.35 0.50 199.50 43.8 17.0 0.0 77.8 Upper/Int Shell I-13 C2613-1 N/A 27.0 0.35 0.49 198.25 43.5 17.0 0.0 104.5 Lower/Int Shell I-14 C2220-1 N/A 27.0 ((0.16)) ((0.64)) ((180.00)) 142.6 8.5 0.0 186.6 Lower/Int Shell I-15 C2220-2 N/A 27.0 ((0.16)) ((0.64)) ((180.00)) 142.6 8.5 0.0 186.6 Lower Shell I-16 A0946-1 N/A 27.0 0.14 0.56 98.20 68.2 17.0 0.0 129.2 Lower Shell I-17 C2193-1 N/A 0.0 0.17 0.50 118.50 82.3 17.0 0.0 116.3 Chemistry Adjustments For 1/4t Filler Initial Chemistry Description Heat No. RTNDT Margin Terms ARTNDT Material RTNDT (°F) Cu (wt%) Ni (wt%) Factor (°F)

(°F) (°F) (°F) (°F)

Limiting Weld - Beltline -- E8018N -65.6 0.10 0.99 134.90 106.9 28.0 12.7 102.8 Chemistry Adjustments For 1/4t Plate Initial Chemistry Description Heat No. RTNDT Margin Terms ARTNDT Location RTNDT (°F) Cu (wt%) Ni (wt%) Factor (°F)

(°F) (°F) (°F) (°F)

Plate Bounding N-2 Nozzle E21VW 40.0 0.18 0.86 141.90 51.2 17.0 0.0 125.2 I-16/I-17 Fluence Data Wall Thickness (in) Fluence at ID Attenuation, 1/4t Fluence at 1/4t Fluence Factor, FF Location Full 1/4t (n/cm2) e-0.24x (n/cm2) f(0.28-0.10 log f)

Upper/Int Shell I-12 5.063 1.266 4.06E+17 0.738 2996E+17 0.219 Upper/Int Shell I-13 5.063 1.266 4.06E+17 0.738 2996E+17 0.219 Lower/Int Shell I-14 5.063 1.266 6.43E+18 0.738 4.746E+18 0.792 Lower/Int Shell I-15 5.063 1.266 6.43E+18 0.738 4.746E+18 0.792 Lower Shell I-16 5.063 1.266 4.46E+18 0.738 3.292E+18 0.694 Lower Shell I-17 5.063 1.266 4.46E+18 0.738 3.292E+18 0.694 Limiting Weld - Beltline 5.063 1.266 6.43E+18 0.738 4.746E+18 0.792 Bounding N-2 Nozzle 5.063 1.266 1.01E+18 0.738 7.454E+17 0.361

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 41 of 84 Reflood Thermal Shock Analysis of the RPV Summary Description The MNGP USAR includes an end-of-life thermal shock analysis performed on the RPV for a design basis LOCA followed by a low-pressure coolant injection. The effects of neutron embrittlement assumed by this thermal shock analysis will change with an increase in the licensed operating period.

This analysis satisfies the criteria of 10 CFR 54.3(a). As such, this analysis is a TLAA.

Analysis For the current operating period, a thermal shock analysis was originally performed on the RPV components. The analysis assumed a design basis LOCA followed by a low-pressure coolant injection accounting for the full effects of neutron embrittlement at the end of life (40 years). The analysis showed that the total maximum vessel irradiation (1 MeV) at the mid-core inside of the vessel to be 2.4 x 1017 n/cm2 which was below the threshold level of any nil-ductility temperature shift for the vessel material. As a result, it was concluded that the irradiation effects on all locations of the RPV could be ignored. However, this analysis only bounded 40 years of operation.

The peak fluence at the RPV wall for the MNGP RPV is 5.17 x 1018 n/cm2 for 54 EFPY of operation (3.90 x 108 MWh through the end of Cycle 22 at 1775 MWt plus 4.76 x 108 MWh at 1880 MWt). Based on this fluence value, the previous analysis is not bounding for the period of extended operation. The original analysis has been superseded by an analysis for BWR-6 RPVs (Reference 7) that is applicable to the MNGP BWR3 RPV. The revised analysis is applicable to MNGP as it uses a bounding main steam line break event, and an RPV thickness similar to the MNGP RPV. This analysis assumes end-of-life material toughness, which in turn depends on end-of-life ART. The critical location for fracture mechanics analysis is at 1/4 of the RPV thickness (from the inside, 1/4T). For this event, the peak stress intensity occurs at approximately 300 seconds after the LOCA.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 42 of 84 Disposition: Validation, 10 CFR 54.21(c)(1)(i)

The current analysis (Reference 7) assumes end-of-life material toughness, which in turn depends on end-of-life ART. The critical location for fracture mechanics analysis is at 1/4 of the vessel thickness (from the inside, 1/4T).

For this event, the peak stress intensity occurs at approximately 300 seconds after the LOCA.

The analysis shows that at 300 seconds into the thermal shock event, the temperature of the vessel wall at 1.5 inches deep (which is 1/4T) is approximately 400°F. For the MNGP vessel, the 1/4T is 1.26 inches. The current analysis is bounding for MNGP for two reasons: (1) the pressure stress (higher for a thinner vessel) is near zero in a thermal shock event, and therefore can be neglected; and (2) the thermal shock event thermal stresses in a 6-inch vessel are greater than those in a 5.06-inch vessel.

Figure 3 of (Reference 7) was used to determine the appropriate parameters for the thinner vessel. Figure 3 demonstrates that 300 seconds into the thermal shock event, the temperature of the vessel wall at 1.26 inches deep is approximately 370°F. The ART values tabulated in Table K3.1-3 list the ARTs for the limiting weld metal of the MNGP RPV. The highest calculated RPV beltline material ART value is 187°F. Using the equation for KIC presented in Appendix A of ASME Section XI (Reference 8) and the maximum ART value, the material reaches upper shelf (a KIC value of 200 ksiin) at 291°F, which is well below the 370°F 1/4T temperature predicted for the thermal shock event at the time of peak stress intensity.

Therefore, the revised analysis is valid for the period of extended operation.

Reflood Thermal Shock Analysis of the RPV Core Shroud Summary Description Radiation embrittlement may affect the ability of RPV internals, particularly the core shroud to withstand a low-pressure coolant injection thermal shock transient. The analysis of core shroud strain due to reflood thermal shock is a TLAA because it is part of the current licensing basis, supports a safety determination, and is based on the calculated lifetime neutron fluence.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 43 of 84 Analysis The RPV core shroud was evaluated for a low-pressure coolant injection reflood thermal shock transient considering the embrittlement effects of 40-year radiation exposure (32 EFPY). The core shroud receives the maximum irradiation on the inside surface opposite the midpoint of the fuel centerline. The total integrated neutron fluence at end of life at the inside surface of the shroud is anticipated to be 2.7 x1020 n/cm2 (greater than 1 MeV). The maximum thermal shock stress in this region will be 155,700 psi equivalent to 0.57% strain. This strain range of 0.57% was calculated at the midpoint of the shroud, the zone of highest neutron irradiation. The calculated strain range of 0.57% represents a considerable margin of safety relative to measured values of percent elongation for annealed Type 304 stainless steel irradiated to 8 x 1021 n/cm2 (greater than 1 MeV). The measured value of percent elongation for stainless steel weld metal is 4% for a temperature of 297°C (567°F) with a neutron fluence of 8 x1021 n/cm2 (greater than 1 MeV), while the average value for base metal at 290°C (554°F) is 20% (Reference 9). Therefore, thermal shock effects on the shroud at the point of highest irradiation level will not jeopardize the proper functioning of the shroud following the design basis accident (DBA) during the current licensed operating period (40 years).

Disposition: Validation, 10 CFR 54.21(c)(1)(i)

As discussed above, core shroud components were evaluated for a reflood thermal shock event, considering the embrittlement effects of lifetime radiation exposure. The analysis includes the most irradiated point on the inner surface of the shroud where the calculated value of fluence for 40-year operating period as below the threshold (3.0 x 1020 n/cm2) for material property changes due to irradiation. However, using the approved fluence methodology discussed above in the section entitled Adjusted Reference Temperature (ART) for RPV Materials Due to Neutron Embrittlement, the 54 EFPY fluence at the most irradiated point on the core shroud was calculated to be 3.84 x 1021 n/cm2.

Because the measured value of elongation bounds the calculated thermal shock strain amplitude of 0.57%, the calculated thermal shock strain at the most irradiated location is acceptable considering the embrittlement effects for a 60-year operating period.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 44 of 84 RPV Thermal Limit Analyses: Operating Pressure - Temperature Limits Summary Description The ART is the value of (Initial RTNDT + RTNDT + margins for uncertainties) at a specific location. Neutron embrittlement increases the ART. Thus, the minimum metal temperature at which an RPV is allowed to be pressurized increases. The ART of the limiting beltline material is used to correct the beltline P-T limits to account for irradiation effects.

10 CFR Part 50, Appendix G requires RPV thermal limit analyses to determine operating pressure-temperature (P-T) limits for boltup, hydrotest, pressure tests and normal operating and anticipated operational occurrences. Operating limits for pressure and temperature are required for three categories of operation: 1) hydrostatic pressure tests and leak tests, referred to as Curve A; 2) non-nuclear heat-up / cooldown and low-level physics tests, referred to as Curve B; and 3) core critical operation, referred to as Curve C. Pressure/temperature limits are developed for three vessel regions: the upper vessel region, the core beltline region, and the lower vessel bottom head region. The calculations associated with generation of the P-T curves satisfy the criteria of 10 CFR 54.3(a). As such, this topic is a TLAA.

Analysis The PTLR contain P-T limit curves for heat up cooldown, and in-service leakage and hydrostatic testing and also limits the maximum rate of change of reactor coolant temperature. The criticality curves provide limits for both heat up and criticality calculated for a 54 EFPY operating period. The current PTLR contains P-T curves developed using the 2004 Edition of the ASME Boiler and Pressure Vessel Code, incorporating the effects of the 1998 power rerate and Code Case N-640. The ART for the period of extended operation (through 54 EFPY) is 186.6F.

Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

NSPM manages the P-T curves in conjunction with surveillance capsule results as part of the BWRVIP Integrated Surveillance Program (BWRVIP-86-A (Reference 5) and BWRVIP-116 (Reference 15),

respectively).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 45 of 84 RPV Circumferential Weld Examination Relief Summary Description Relief from RPV circumferential weld examination requirements under Generic Letter (GL) 98-05 is based on probabilistic assessments that predict an acceptable probability of failure per reactor operating year. The analysis is based on RPV metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period.

MNGP has received this relief for the remaining 40 year licensed operating period. The circumferential weld examination relief analysis meets the requirements of 10 CFR 54.3(a) (Reference 2). As such, they are a TLAA.

Analysis MNGP received NRC approval for a technical alternative that eliminated the RPV circumferential shell weld inspections for the current license term. The basis for this relief request was an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIP-05 and the extent of neutron embrittlement. The anticipated changes in metallurgical conditions expected over the extended licensed operating period require an additional analysis for 54 EFPY and approval by the NRC to extend this relief request.

Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

The USNRC evaluation of BWRVIP-05 used the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate the RPV shell weld failure probabilities (Reference 10). Three key assumptions of the PFM analysis are: 1) the neutron fluence was the estimated end-of-life mean fluence; 2) the chemistry values are mean values based on vessel types; and 3) the potential for beyond-design-basis events is considered. Table K3.1-4 provides a comparison of the MNGP RPV limiting circumferential weld parameters to those used in the NRC analysis for the first two key assumptions. Data provided in Table K3.1-4 was supplied from Tables 2.6-4 and 2.6-5 of the Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report.

For MNGP, the chemistry values are the same as those used in the NRC analysis, however, the chemistry factor is higher due to an adjustment to reflect the results from two surveillance capsules. The value of fluence is lower than that used in the NRC analysis. As a result, the shift in reference temperature is lower than the 64 EFPY shift from the NRC analysis. In addition, the unirradiated reference temperature is essentially the same.

The combination of unirradiated reference temperature (RTNDT(U)) and shift (RTNDT w/o margin) yields an ART that is lower than the NRC mean analysis value.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 46 of 84 Therefore, the RPV shell weld embrittlement due to fluence has a negligible effect on the probabilities of RPV shell weld failure. The Mean RTNDT value at 54 EFPY is bounded by the 64 EFPY Mean RTNDT provided by the NRC.

Although a conditional failure probability has not been calculated, the fact that the MNGP values at the end of license are less than the 64 EFPY value provided by the NRC leads to the conclusion that the MNGP RPV conditional failure probability is bounded by the NRC analysis.

The procedures and training used to limit reactor pressure vessel cold over-pressure events will be the same as those approved by the NRC when MNGP requested approval of the BWRVIP-05 technical alternative for the term of the current operating license. A request for extension for the 60-year extended operating period will be submitted to the NRC prior to the period of extended operation.

MNGP Update: A 10 CFR 50.55a Request (relief request) was submitted to the NRC to continue the alternative that eliminates the RPV circumferential shell weld inspections for the period of extended operation. The relief is based on the requirements of BWRVIP-05 and BWRVIP-74, and is consistent with NRC Generic Letter 98-05. The NRC approved the relief for the remainder of the license renewal extended period of operation.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 47 of 84 Table K3.1-4 Effects of Irradiation on RPV Circumferential Weld Properties for MNGP CB&I 64 EFPY MNGP 54 EFPY Group (Reference 10) (Reference 27)

Cu (%) 0.10 0.10 Ni (%) 0.99 0.99 CF 134.9 134.9 Fluence at clad/weld interface (1019 n/cm2) 1.02 0.64 RTNDT w/o margin (F) 135.6 106.9 RTNDT(U) (F) -65 -65.6 Mean RTNDT (F) 70.6 41.3 P (F/E) NRC1 1.78 x 10-5 2 P (F/E) BWRVIP - -

RPV Axial Weld Failure Probability Summary Description The BWRVIP recommendations for inspection of RPV shell welds (Reference 11) contain generic analyses supporting an NRC SER (Reference 10) conclusion that the generic-plant axial weld failure rate is no more than 5 x 10-6 per reactor year. BWRVIP-05 showed that this axial weld failure rate of 5 x 10-6 per reactor year is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described above in the section entitled RPV Circumferential Weld Examination Relief.

1 P (F/E) stands for probability of a failure event.

2 Although a conditional failure probability has not been calculated, the fact that the MNGP values at the end of license are less than the 64 EFPY value provided by the NRC leads to the conclusion that the MNGP RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements of Reference 4.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 48 of 84 MNGP received relief from the circumferential weld inspections for the remaining 40 year licensed operating period. The axial weld failure probability analysis meets the requirements of 10 CFR 54.3(a) (Reference 2). As such, it is a TLAA.

Analysis As stated above in the section entitled RPV Circumferential Weld Examination Relief, MNGP received NRC approval for a technical alternative that eliminated the RPV circumferential shell weld inspections for the current license term. The basis for this relief request was an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIP-05 and the extent of neutron embrittlement. The NRC SER associated with BWRVIP-05 (Reference 10) concluded that the RPV failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is less than 5 x 10-6 per reactor year. This failure frequency is dependent upon given assumptions of flaw density, distribution, and location. The failure frequency also assumes that essentially 100% of the RPV axial welds will be inspected. The anticipated changes in metallurgical conditions expected over the extended licensed operating period require an additional analysis for 54 EFPY and approval by the NRC to extend the RPV circumferential weld inspection relief request.

MNGP Update: A 10 CFR 50.55a Request (relief request) was submitted to the NRC to continue the alternative that eliminates the RPV circumferential shell weld inspections for the period of extended operation. The relief is based on the requirements of BWRVIP-05 and BWRVIP-74, and is consistent with NRC Generic Letter 98-05. The NRC approved the relief for the remainder of the license renewal extended period of operation.

Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

Table K3.1-5 compares the limiting axial weld 54 EFPY properties for MNGP against the values taken from Table 2.6-5 found in the NRC SER for BWRVIP-05 and associated supplement to the SER (Reference 12). The SER supplement required the limiting axial weld to be compared with data found in Table 3 of the document. For MNGP, the comparison was made to the 'Mod 2' plant information. The supplemental SER stated that the 'Mod 2' calculations most closely match the 5 x 10-6 RPV failure frequency.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 49 of 84 For MNGP, the fluence value is greater than that used in the NRC analysis.

However, the weld material has a significantly lower copper value (0.10 vs.

0.219 used in the NRC analysis); the nickel values are essentially the same as those used in the NRC analysis. As a result, the value of RTNDT is lower than the NRC analysis. In addition, the unirradiated RTNDT was significantly lower (-65.6°F vs. -2°F used in the NRC analysis). The MNGP limiting weld 54 EFPY Mean RTNDT value is within the limits of the values assumed in the analysis performed by the NRC staff in the March 7, 2000, BWRVIP-05 SER supplement and the 64 EFPY limits and values obtained from Table 2.6.5 of the SER. Therefore, the probability of failure for the axial welds is bounded by the NRC evaluation.

Table K3.1-5 Effects for Irradiation on RPV Axial Weld Properties for MNGP Mod 2 MNGP 54 EFPY Value (Reference 4) (Reference 27)

Cu (%) 0.219 0.10 Ni (%) 0.996 0.99 CF 134.9 Fluence x 1019 (n/cm2) 0.1483 0.64 RTNDT (F) 116 106.9 RTNDT(U) (F) -2 -65.6 Mean RTNDT (F) 114 41.3 P (F/E) NRC 5.02 x 10-6 4 3

Peak Axial Fluence 4

Although a conditional failure probability has not been calculated, the fact that the MNGP values at the end of license are less than the Mod 2 value provided by the NRC leads to the conclusion that the MNGP RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements of Reference 4.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 50 of 84 K3.2 Metal Fatigue of the RPV and Internals, and Reactor Coolant Pressure Boundary Piping and Components A cyclically loaded metal component may fail because of fatigue even though the cyclic stresses are considerably less than the static design limit. Some design codes such as the ASME Boiler and Pressure Vessel Code and the ANSI piping codes contain explicit metal fatigue calculations or design limits. Cyclic or fatigue design of other components may not be to these codes, but may use similar methods. These analyses, calculations and designs to cycle count limits or to fatigue usage factor limits may be TLAAs.

Fatigue analyses are presented in the following groupings:

• RPV Fatigue Analyses

• RPV Internals Fatigue Analysis NUREG-1801 identifies numerous fatigue related aging effects that require evaluation as possible TLAAs in accordance with 10 CFR 54.21(c). Each of these is summarized in NUREG-1800.

RPV Fatigue Analyses Summary Description RPV fatigue analyses were performed for the vessel support skirt, shell, upper and lower heads, closure flanges, nozzles and penetrations, nozzle safe ends, and closure studs. The end-of-40-year license fatigue usage was determined for the normal and upset pressure and thermal cycle events.

Subsequent to the original stress analyses, several hardware changes, operational changes (such as the 1998 power rerate), and/or stress analysis revisions have affected the usage factors.

Calculation of fatigue usage factors is part of the current licensing basis and is used to support safety determinations. The RPV fatigue analyses are TLAAs.

Analysis The original RPV stress report included a fatigue analysis for the RPV components based on a set of design basis duty cycles. The original 40-year analyses demonstrated that the CUFs for the critical components would remain below the ASME Code Section III allowable value of 1.0.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 51 of 84 A reanalysis was performed for RPV CUF values as a part of the 1998 power rerate implementation at MNGP. For power rerate implementation, only components in which the original and power rerate modification stress report CUF values are greater than 0.5 required reanalysis. Subsequent to the original and modification analyses, a fatigue monitoring program was developed and revised fatigue usage values were determined. These fatigue usage values considered actual thermal cycle experience through September 30, 2004. The resulting fatigue CUF values determined for the monitoring program and power rerate supersede the values determined in the original and modification RPV analyses. The CUF values as of September 30, 2004 and 60-year CUF values are listed in Table K3.2-1. As part of the MNGP fatigue monitoring program, a review of plant transients has been conducted for the period of September 30, 2004 through December 31, 2006. This review confirmed that the cycle projections and CUF values identified in Table K3.2-1 are still valid.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 52 of 84 Table K3.2-1 Fatigue Evaluation Results for Limiting Components Computed Computed Fatigue Monitoring Fatigue Usage Usage Recommended Component Factor Factor 60- by NUREG/

(through Year CR-6260 9/30/2004)

License Recirculation Outlet Nozzle 0.010 0.015 Yes Recirculation Inlet Nozzle 0.145 0.220 Yes Steam Outlet Nozzle 0.124 0.187 No Feedwater Nozzle 0.328 0.597 Yes Core Spray Nozzle 0.233 0.645 Yes Core Support Structure 0.039 0.058 No Bottom Head and Support Skirt 0.206 0.293 Yes Control Rod Drive Penetrations 0.179 0.288 No Vessel Closure Bolts 0.340 0.554 No Refueling Bellows Skirt 0.502 0.829 No The 60-year results incorporate current fatigue monitoring program cycles accumulated through December 31, 2006. Cycle counting includes those cycles identified in MNGP USAR Table 4.2-1, which identifies the following transient cycles:

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 53 of 84 MNGP Transient Cycles No. of Design Projected Transient Type Cycles (USAR to 2030 Table 4.2-1)

Bolt Up/Unbolt 120 44 Startup/Shutdown @ 100oF/hr 289 207 Scrams 270 165 Design Hydrostatic Test @ 1250 psig 130 67 Reactor Overpressure @ 1375 psig 1 0 Hydrostatic Test to 1560 psig 3 2 Rapid Blowdown 1 0 Liquid Poison Flow @ 80oF 10 0 Feedwater Heater Bypass 70 0 Loss of Feedwater Heater 10 0 Loss of Feedwater Pumps 30 0 Improper Start of Shutdown Recirc Loop 10 8 It should be noted that not all cycles apply to all locations evaluated, and that the number of design cycles identified above represent design values, not the maximum allowable number of transients.

The original code analysis of the reactor vessel included fatigue analysis of the control rod drive hydraulic system return line nozzles. After several years of operation, it was discovered that the control rod drive hydraulic system return line nozzles were subject to cracking caused by a number of factors including rapid thermal cycling (Reference 13). Consequently, the control rod drive hydraulic system return line nozzles were capped and removed from service. As such, they are no longer subject to rapid thermal cycling.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 54 of 84 Disposition: Revision and Aging Management, 10 CFR 54.21(c)(1) (ii) and (iii)

For the period of extended operation, the fatigue usage factors for the limiting components have been re-evaluated. No MNGP component exceeded the ASME Code allowable for the 60-year license. The results of the evaluation are shown in Table K3.2-1.

As stated in Chapter IV.A1 of the Generic Aging Lessons Learned (GALL)

(NUREG-1801), environmental fatigue issues must be considered for Class 1 components. Chapter 4.3 (Metal Fatigue) of NUREG-1800 states that an aging management program consistent with Chapter X.M1 of the GALL is an acceptable method for management of metal fatigue for the period of extended operation. The current fatigue monitoring program tracks CUFs through cycle-based fatigue (CBF) monitoring.

CBF monitoring consists of a two-step process: (a) cycle counting, and (b)

CUF computation based on the counted cycles. The cycle counting counts each transient that is defined in the plant-licensing basis based upon the mechanistic process or sequence of events experienced by the plant as determined from monitored plant instruments. The approach is conservative because it assumes each actual transient has a severity equal to that assumed in the design basis. Transients are identified and implemented into the aging management program. CUF computation calculates fatigue directly from counted transients and parameters for the monitored components. CUF is computed via a design-basis fatigue calculation where the numbers of cycles are substituted for assumed design basis number of cycles.

The current fatigue monitoring program includes 10 components listed in Table K3.2-1. With environmental fatigue considered (see Section K3.7), this program meets the recommendations of Chapter X.M1 of the GALL for the period of extended operation. This is consistent with the components listed in NUREG/CR-6260 (Reference 14), and the recommendations of Chapter X.M1 of the GALL.

Fatigue Analysis of RPV Internals Summary Description Fatigue analysis of the RPV internals was performed using the ASME Boiler and Pressure Vessel Code,Section III, as a guide. The most significant fatigue loading occurs at the jet pump diffuser to baffle plate weld location.

The original 40-year calculation showed a CUF of ~0.33, less than the ASME allowable of 1.0 (Reference 6). Because this analysis used a number of cycles for a 40-year life, it is a TLAA.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 55 of 84 Analysis The events analyzed included: (1) Normal startup and shutdown; (2)

Improper start of a recirculation loop; and (3) DBA. The fatigue evaluation determined that peak strains occurred as a result of the improper recirculation loop startup transient and the point in the time of the DBA flooding (Low Pressure Coolant Injection (LPCI)) when the shroud and shroud support plate through-wall gradients are at a maximum. None of the other events analyzed contributed significantly to fatigue usage. The 40-year CUF for this location was determined to be ~0.33, i.e., less than the ASME allowable of 1.0.

Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

Because the original fatigue analysis used a number of cycles for a 40-year design life, the calculation was revised for a 60-year life by scaling up the number of cycles by 1.5, except for the DBA transient. The resultant fatigue usage was calculated to be ~0.5, which is less than the ASME Code allowable of 1.0. Therefore, the fatigue usage of the RPV internal components is acceptable for the period of extended operation.

K3.3 ASME Section III Class 1 Reactor Coolant Pressure Boundary (RCPB) Piping Fatigue Analysis Summary Description MNGP piping systems were originally designed in accordance with ASA B31.1 and USAS B31.1.0, which did not require that an explicit fatigue analysis be performed.

Reconciliation for the use of later editions of construction codes for modification to or replacement of piping and components has been performed in accordance with Section IWA-7210(c),Section XI of the ASME Code. The governing code for design, materials, fabrication and erection of piping, piping components, and pipe support modifications or replacements is ANSI B31.1, 1977 Edition including Addenda up to and including the Winter of 1978.

Portions of Class 1 systems such as the Reactor Recirculation, Core Spray and RHR inside drywell were required to be analyzed for fatigue in accordance with the ASME Code Section III for Nuclear Class I piping. The implementation of these requirements at MNGP were for the purpose of attaining a higher quality level and provide more detailed analysis to confirm protection of the reactor coolant system integrity.

The analyses demonstrate that the 40 year cumulative usage factors (CUF) for the limiting components in all effected systems are below the ASME Code Section III allowable value of 1.0. Because these analyses are based on cycles postulated to occur in the current 40 year design life, they are TLAAs.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 56 of 84 Analysis With the exception of the torus attached piping and safety relief discharge line piping which were evaluated as part of the Mark I program New Loads program, the only piping that has been explicitly analyzed for fatigue are portions of the recirculation system piping, RHR piping, and core spray piping systems. These systems were all modified under the Generic Letter (GL) 88-01 IGSCC inspection and mitigation program.

This piping was originally designed in accordance with USAS B31.1 and modifications were analyzed to ASME Section III Class 1 rules. The ASME Code limit for fatigue is 1.0. The limiting fatigue usages for these systems are shown in Table K3.3-1.

Table K3.3-1 MNGP Fatigue Monitoring Locations for RCPB Class 1 Piping 40 Year Cumulative Location Fatigue Usage Factor Recirculation Equalizer Line Branch Connection 0.8514 RHR Return Loop B Tapered Transition 0.8875 Core Spray Valve Joint 0.6466 For fatigue analyses, the change in stress produced by transients are compared to allowable limits. For a given stress range, the ASME code allows a maximum number of cycles. In a fatigue analysis the actual or design assumed number of cycles is compared to the allowed maximum, and this ratio is summed for all significant transients experienced by the component. The summation, or usage factor, must be less than or equal to 1.0 to be acceptable.

The fatigue analyses for these systems were evaluated using a bounding set of assumed thermal cycles that may occur over the life of the plant (40 years).

These conservative evaluations resulted in fatigue usage values that are acceptable (i.e. less than 1.0) however, with the exception of the core spray piping there is not sufficient margin to extrapolate by a ratio of 1.5 with acceptable results.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 57 of 84 Cycle based counting consists of periodically counting the relevant cycles and calculating the cumulative usage factor (CUF). This process is also conservative due to the fact that all transients within a group are assumed to be equal in severity and correspond to the maximum cycle thermal limits specified in the design. Based on the number of cycles experienced at MNGP through September 2004 (confirmed by review of cycles experienced October 2004 through December 2006), the maximum fatigue usages identified in Table K3.3-1 for the Recirculation and RHR piping systems are expected not to exceed 0.90 at the end of sixty (60) years of plant operation.

Disposition: Validation, 10 CFR 54.21(c)(1)(i) and Aging Management 10 CFR 54.21(c)(1)(iii)

The limiting location for RCPB core spray piping is less than 0.65. Consequently, current analyses are validated for the period of extended operation by:

Umax,40 <0.65, x 60/40 = Umax,60 = 0.975 < 1.0 The limiting locations for the recirculation and RHR piping are less than 0.90 taking into account actual cycles accumulated through 2002 and projecting those cycles to 60 years. The MNGP cycle based fatigue monitoring system manages this aging mechanism to ensure that fatigue does not exceed the allowable limit of 1.0.

K3.4 RCPB Section III Class 2 AND 3, ASA B31.1 AND, USAS B31.1 Piping and Components Summary Description MNGP piping systems were originally designed in accordance with ASA B31.1 and USAS B31.1.0, which did not require that an explicit fatigue analysis be performed.

Reconciliation for the use of later editions of construction codes for modification to or replacement of piping and components has been performed in accordance with Section IWA-7210(c),Section XI of the ASME Code. The governing code for design, materials, fabrication and erection of piping, piping components, and pipe support modifications or replacements is ANSI B31.1, 1977 Edition including Addenda up to and including the Winter of 1978.

The codes and standards which MNGP was designed and constructed to did not include fatigue analyses for piping, component supports or component connections and anchors. The only exceptions are some ASME Class MC containment piping support and penetration analyses for New Loads (Section K3.8), and RCPB piping discussed in the preceding section.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 58 of 84 Analysis Although the code of construction did not invoke fatigue analyses, a stress range reduction factor which is applied to the allowable stress range for expansion stresses is required to account for cyclic thermal conditions. The allowable secondary stress range is 1.0 SA for 7,000 equivalent full temperature thermal cycles or less and is incrementally reduced to 0.5 SA for greater than 100,000 cycles. With the exception of piping described in Section K3.3 and Section K3.8, MNGP piping analyses incorporated stress range reduction factors for a finite number of thermal cycles in lieu of fatigue analyses.

Disposition: Validation, 10 CFR 54.21(c)(1)(i)

An estimate of the number of thermal cycles experienced by these piping systems can be conservatively approximated by the maximum number of thermal cycles used in reactor nozzle fatigue analyses. For MNGP the bounding number of cycles used for the qualification of a vessel nozzle is 1,500 for the feedwater nozzle. The maximum number of cycles projected through the extended period of operation is therefore, 1.5 times 1,500 (2,250). This conservative amount of full range cycles is significantly less than the 7000 cycle limit, consequently existing analyses are valid through the extended term of operation.

K3.5 Irradiation Assisted Stress Corrosion Cracking Summary Description Austenitic stainless steel RPV internal components exposed to a neutron fluence greater than 5 x 1020 n/cm2 (E > 1 MeV) are susceptible to irradiation assisted stress corrosion cracking (IASCC) in the BWR environment. As described in the SER to BWRVIP-26, IASCC of RPV internals is a TLAA.

Analysis Fluence calculations have been performed for the RPV and internals, including the effects of an extended power uprate (2004 MWt). Three components have been identified as being susceptible to IASCC for the period of extended operation: (1) Top Guide, (2) Shroud, and (3) Incore Instrumentation Dry Tubes and Guide Tubes.

Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)

The top guide, shroud, and incore instrumentation dry tubes and guide tubes are susceptible to IASCC. The aging effect associated with IASCC (crack initiation and growth) will require aging management. All three components (top guide, shroud, and incore instrumentation dry tubes and guide tubes) have been evaluated by the BWRVIP, as described in the Inspection and Evaluation Guidelines for each component: BWRVIP-26 (Top Guide), BWRVIP-76 (Shroud),

and BWRVIP-47 (incore instrumentation dry tubes and guide tubes). BWRVIP recommendations are implemented at MNGP by the Water Chemistry and the In-Service Inspection Programs.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 59 of 84 K3.6 Stress Relaxation of Rim Holddown Bolts Summary Description As described in the SER to BWRVIP-25, plants must consider relaxation of the rim-hold-down bolts as a TLAA issue. Because MNGP has not installed core plate wedges, the loss of preload must be considered in the TLAA evaluation.

Analysis The core plate hold-down bolts connect the core plate to the core shroud. These bolts are subject to stress relaxation due to thermal and irradiation effects. For the 40-year lifetime, the BWRVIP concluded that all rim hold-down bolts would maintain some preload throughout the life of the plant.

Disposition: Revision 10 CFR 54.21(c)(1)(ii)

For MNGP the projected loss of preload at the end of the period of extended operation is 8% based on the MNGP design and a neutron fluence of 2.2 x 1019 n/cm2 (E > 1.0 Mev). This neutron fluence corresponds to the maximum fluence applicable to the bolts at the end of the period of extended operation, although many bolts experience lower fluence due to there specific azimuthal location. The fluence calculation was performed using methodology in accordance with the guidance provided in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The MNGP core plate rim hold-down bolt evaluation demonstrates that the mean axial and bending stresses, considering hold-down bolt stress relaxation, are bounded by the BWRVIP-25 analysis results and/or the ASME Code allowable limits.

K3.7 Effects of Reactor Coolant Environment Summary Description Generic Safety Issue (GSI)-190 was identified by the NRC because of concerns about the effects of reactor water environments on the fatigue life of components and piping during the period of extended operation. GSI-190 was closed in December of 1999, and concluded that environmental effects have a negligible impact on core damage frequency, and as such, no generic regulatory action is required. However, as part of the closure of GSI-190, the NRC concluded that licensees who apply for license renewal should address the effects of coolant environment on component fatigue life as part of their aging management programs.

Fatigue calculations that include consideration of environmental effects to establish cumulative usage factors can be treated as TLAAs under 10 CFR Part 54 or they could be used to establish the need for an aging management program.

To qualify as a TLAA, the analysis must satisfy all (6) criteria defined in 10 CFR 54.3. Failure to satisfy any one of these criteria eliminates the analysis from further consideration as a TLAA.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 60 of 84 Fatigue design for MNGP has been determined to be a TLAA, even though the design limits are based on cycles rather than an explicit time period. Reactor water environmental effects, however, are not included in the MNGP current licensing basis (CLB). Consequently, the criterion of 10 CFR 54.3(a)(6) is not satisfied. Nevertheless, environmental effects on Class 1 component fatigue have been evaluated to determine if any additional actions are required for the extended period of operation.

Analysis The NRC staff assessed the impact of reactor water environment on fatigue life at high fatigue locations and presented the results in NUREG/CR-6260, Application of NUREG/CR-5999, Interim Fatigue Curves for Selected Nuclear Power Plant Components, in March of 1995. Methodology for the determination of environmental correction factors to be applied to the fatigue analyses for carbon and low-alloy steels is contained in NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels.

Methodology for environmental fatigue factors for austenitic stainless steels is contained in NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue Design of Austenitic Stainless Steels.

As a part of the NRCs Fatigue Action Plan, incorporation of environmental fatigue effects originally involved a reduced set of fatigue design curves, such as those proposed by Argonne National Laboratory (ANL) in NUREG/CR-5999. As a part of the effort to close GSI-166 (later GSI-190) for operating nuclear power plants during the current 40-year licensing term, Idaho National Engineering Laboratory (INEL) evaluated fatigue-sensitive component locations at plants designed by all four U. S. nuclear steam supply system (NSSS) vendors. The ANL fatigue curves were used by INEL to recalculate the cumulative usage factors (CUFs) for fatigue-sensitive component locations in early and late vintage Combustion Engineering (CE) pressurized water reactors (PWRs), early and late vintage Westinghouse PWRs, early and late vintage General Electric (GE) boiling water reactors (BWRs), and Babcock & Wilcox Company (B&W) PWRs. The results of the INEL calculations were published in NUREG/CR-6260 (Reference 14). The INEL calculations took advantage of conservatisms present in governing ASME Code fatigue calculations, including the numbers of actual plant transients relative to the numbers of design-basis transients, but did not recalculate stress ranges based on actual plant transient profiles. The BWR calculations, especially the early-vintage GE BWR calculations, are directly relevant to MNGP.

In order to comply with the requirements, MNGP has evaluated the locations specified in NUREG/CR-6260 for the older vintage BWR plants. These locations consist of:

• Reactor Vessel (Lower Head to Shell Transition)

• Feedwater Nozzle

• Recirculation System (Vessel Nozzles and RHR Return Line Tee)

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• Core Spray System (Nozzle/Safe End)

• Residual Heat Removal Piping (Tapered Transition)

• Limiting Feedwater Piping Location For each location, detailed environmental fatigue calculations have been performed using Fen relationships for carbon and low-alloy steel locations (NUREG/CR-6583) and stainless steel locations (NUREG/CR-5704). The calculations incorporate Fen methodology to determine a multiplier on the cumulative usage factor (CUF) so that environmental effects can be assessed. As can be seen in Table K3.7-1, all locations are acceptable through the extended term of operation due to the fact that all CUFs remain below the acceptance criteria of 1.0.

Table K3.7-1 Summary of Environmental Fatigue Usage Factors for MNGP Usage Factor Location Component Material (Uenv)

Reactor Vessel Shell Carbon Steel 0.748 Feedwater Nozzle Safe End Carbon Steel 0.872 Recirculation Inlet Nozzle Safe End Stainless Steel 0.749 Core Spray Nozzle Safe End Carbon Steel 0.268 Recirculation Piping RHR Tee Stainless Steel 0.864 Feedwater Piping FWTR/RCIC Tee Carbon Steel 0.673 Disposition: Revision 10 CFR 54.21(c)(1)(ii)

The cumulative usage factors for all locations, when conservatively re-evaluated to include environmental effects, remains below 1.0. Although, based on a projection of experienced cycles, these locations have been shown to be acceptable through the period of extended operation, the MNGP thermal fatigue monitoring program periodically reviews and updates fatigue analyses to ensure continued compliance with fatigue acceptance criteria.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 62 of 84 K3.8 Fatigue Analyses of the Primary Containment, Attached Piping, and Components The MNGP primary containment was designed in accordance with the ASME Code,Section III, 1965 Edition with addenda up to and including Winter of 1965.

Subsequently, during large scale testing for the Mark III containment system and the in-plant testing for Mark I primary containment systems, new suppression chamber hydrodynamic loads were identified. These new loads are related to the loss-of-coolant-accident (LOCA) scenario and safety relief valve (SRV) operation.

Containment fatigue analyses are provided for the following groups:

• Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers

• Fatigue Analysis of the SRV Discharge Piping Inside the Suppression Chamber and Internal Structures

• Fatigue Analysis of Suppression Chamber External Piping and Penetrations

• Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analysis

• Primary Containment Process Penetration Bellows Fatigue Analyses Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers Summary Description New hydrodynamic loads were identified subsequent to the original design for the containment suppression chamber vents. These loads result from blowdown into the suppression chamber during a postulated LOCA and during SRV operation for plant transients. The results of analyses of these effects are presented in the MNGP USAR. Consequently, these analyses are TLAAs.

Analysis Analysis of the suppression chamber, vent system and downcomers (Reference 17) identified that the vent header-downcomer intersection and the torus shell were limiting in terms of fatigue usage. Fatigue usages for all other locations were found to be less than 0.015. The calculated values for the vent header-downcomer intersection and the torus shell were 0.684 and 0.66 respectively. Subsequent to that evaluation, all locations were re-evaluated for the effects of power rerate implemented in 1998. It was estimated that power rerate conditions could result in an increase in the number of SRV cycles experienced due to higher steaming rates at increased power levels. The number of cycles was estimated to increase by 14 percent coincident with the increase to 1775 MWt (from 1670 MWt) and by 26 percent due to an increase to 1880 MWt.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 63 of 84 The revised fatigue evaluation conservatively estimated the fatigue usage of the vent header-downcomer intersection as 0.862 (1.26 x 0.684). The revised maximum fatigue for the torus shell was similarly calculated to be 0.98, using increased SRV actuations postulated for rerate conditions and applicable event combinations.

Disposition: Validation, 10 CFR 54.21(c)(1)(i) and Aging Management, 10 CFR 54.21(c)(1)(iii)

All locations with the exception of the vent header-downcomer intersection and the torus shell have reported 40 year fatigue usage factors of less than 0.2. Consequently, those locations are validated by review of the current analyses (e.g. Umax,40 < 0.20, x 60/40 = Umax,60 = 0.30<<1.0).

Since only the SRV load cases contribute to fatigue during normal operation, operation may continue until the contribution from SRV discharges has not exceeded the conservative design values used in the evaluation.

The MNGP cycle based fatigue monitoring program includes periodic counting of the SRV cycles, comparing the total number of experienced SRV cycles to the design basis number of cycles and, confirming that the fatigue usage will remain below the acceptance criteria of 1.0 or identifying when the limit is likely to be exceeded such that adequate corrective measures can be implemented. As of December 31, 2006 the total number of normal operation SRV lifts experienced at the MNGP was 514 and the design basis is 934. Extrapolation of current SRV lifts results in an conservative estimate due to the fact that counted lifts do not differentiate the operating condition at which the lift was experienced (e.g., power level), the design value of 934 postulates that all SRVs lifts occur in the same suppression chamber bay and, the rate of SRV challenges experienced in the first 7 years of operation is significantly higher than subsequently experienced. Without consideration for these conservatisms, 342 additional challenges can be expected throughout the 60 year extended operating period. This would result in a 60 year SRV total of 856.

All applicable plant cycles are currently monitored to ensure that the cumulative usage factors remains below 1.0 for the limiting components. In the unlikely event that fatigue usage is predicted to exceed 1.0 prior to 60 years of operation, appropriate corrective action will be taken in accordance with the MNGP Corrective Action Program.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 64 of 84 Fatigue Analysis of the SRV Piping Inside the Suppression Chamber and Internal Structures Summary Description The Reactor Pressure Relief System includes safety/relief valves (SRVs) located on the main steam lines within the drywell between the reactor vessel and the first isolation valve. The SRVs, which discharge to the suppression pool, provide two main protective functions:

• Overpressure relief - The valves open to limit the pressure rise in the reactor.

• Depressurization - The valves are opened to depressurize the reactor.

The Plant Unique Analysis Report (Reference 18) describes the fatigue analysis of the SRV discharge lines. These analyses assume a limited number of SRV actuations throughout the 40 year life of MNGP and are therefore TLAAs.

Torus internal structures (i.e., catwalk and monorail) are Service Level E structures. Consequently, no fatigue evaluation is required to demonstrate acceptability of these structures.

Analysis The criteria presented in Volume 5 of the MNGP PUAR (Reference 18) describes the evaluation of the SRVDL piping system. The evaluation included the effects of LOCA related loads and SRV discharge related loads.

LOCA and SRV discharge loads were formulated using procedures and test results which included the effects of plant unique geometry and operating parameters contained in the Plant Unique Load Definition (PULD) report (Reference 19). The analysis also considered the interaction effects of the vent system and the suppression chamber.

Per (Reference 18), the critical location for fatigue usage is the SRV piping at the elbow adjacent to the elbow support beam junction. The fatigue usage for this location was calculated to be 0.309.

Subsequent to that evaluation, this location was reevaluated for the effects of power rerate implemented in 1998. It was estimated that power rerate conditions could result in an increase in the number of SRV cycles experienced due to higher steaming rates at increased power levels. The number of cycles was estimated to increase by 14 percent coincident with the increase to 1775 MWt (from 1670 MWt) and by 26 percent due to an increase to 1880 MWt. Conservatively using the 1880 MWt SRV factor, an increase to 0.389 was calculated.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 65 of 84 Disposition: Validation, 10 CFR 54.21(c)(1)(i)

The limiting location for the SRV piping is less than 0.40. Current analyses are validated by:

Umax,40 < 0.40, x 60/40 = Umax,60 <0.60 < 1.0 This increase in service life does not significantly effect SRV discharge piping fatigue usage. Consequently, the current calculation is validated for the period of extended operation.

Fatigue Analysis of Suppression Chamber External Piping and Penetrations Summary Description These analyses include the large and small bore torus attached piping (TAP), suppression chamber penetrations and the ECCS suction header.

Fatigue analyses were completed that were based on cycles postulated to occur within the 40 year operating life of the plant. Therefore these calculations are TLAAs.

Analysis Rigorous analytical techniques were used to evaluate the effects of LOCA related and SRV discharge loads as defined in the NRC's Safety Evaluation Report NUREG-0661 and in the Mark I Containment Load Definition Report (Reference 20). These techniques included detailed analytical models and refined methods for computing the dynamic response of the TAP systems which included consideration of the interaction effects of each piping system and the suppression chamber.

The results of the TAP structural analysis for each load type were used to evaluate load combinations for the piping and penetrations in accordance with NUREG-0661 and the Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide (PUAGG). The analysis results were compared with the acceptance limits specified in the PUAAG and the applicable sections of the ASME Code for Class 2 piping and for Class MC components.

Fatigue effects were specifically addressed for the suppression chamber penetrations and the suction header, whereas the evaluation for the piping was generically addressed for all Mark I plants by the Mark I Owners' Group.

Analyses documented in this report identify cumulative usage factors for the Mark I plants of less than 0.5. The generic fatigue evaluation included 36 piping systems from 15 plants. Stress results for the most limiting piping systems and locations were selected for each plant. Thus, the reported usage factors are representative of the most limiting location within the data for the plant group. For MNGP, the SRV discharge piping was identified as the limiting location. The SRV discharge piping was re-evaluated for the effects of power rerate, which was implemented in 1998. It was estimated that power rerate conditions could result in an increase in the number of

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 66 of 84 SRV cycles experienced due to higher steaming rates at increased power levels. The number of cycles was estimated to increase by 14 percent coincident with the increase to 1775 MWt (from 1670 MWt) and by 26 percent due to an increase to 1880 MWt. Conservatively using the 1880 MWt SRV factor, an increase to 0.389 was calculated.

The TAP penetration fatigue usage was conservatively evaluated for the effects of power rerate by increasing the SRV cycles by a factor of 1.26 to correspond to a power level of 1880 MWt (the actual rerate power level was 1775 MWt, which corresponds to a 1.14 SRV factor). This conservative application, in addition to the bounding analysis, confirmed that fatigue usage for the TAP penetrations would remain below 1.0 (0.985) based on cycles anticipated to occur during the 40 year operating life of MNGP.

Disposition: Validation, 10 CFR 54.21(c)(1)(i) and Aging Management, 10 CFR 54.21(c)(1)(iii)

The limiting location for TAP is less than 0.40. Current analyses are validated by:

Umax,40 < 0.40, x 60/40 = Umax,60 <0.60 < 1.0 This increase in service life does not significantly effect TAP fatigue usage.

Consequently, the current calculation is validated for the period of extended operation.

Conversely, although TAP penetration fatigue usage has been conservatively validated for 40 years of operation there is not sufficient margin to project additional cycles for a 60 year extended term of operation and remain below the acceptance criteria of 1.0.

Since SRV load cases are the primary contributor to fatigue during normal operation, operation may continue until the contribution from SRV discharges has not exceeded the conservative design values used in the evaluation.

The MNGP cycle based fatigue monitoring includes periodic counting of the SRV cycles. The SRV cycles are compared to the design basis number of cycles to confirm that the fatigue usage will remain below the acceptance criteria of 1.0 and to provide timely identification of when the limit may be exceeded such that adequate corrective measures can be enacted. As of December 31, 2006 the total number of SRV lifts experienced at the MNGP was 514. Projecting this rate of SRV lifts throughout 60 years of operation indicates that the fatigue usage will remain below 1.0 for the period of extended operation.

All applicable plant cycles are currently monitored to ensure that the cumulative usage factors remains below 1.0 for the limiting components. In the unlikely event that fatigue usage is predicted to exceed 1.0 prior to 60 years of operation, appropriate corrective action will be taken in accordance with the MNGP Corrective Action Program.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 67 of 84 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analysis Summary Description The drywell-to-suppression chamber vent line bellows are included in the Mark I Containment Long Term Program plant-unique analysis. A fatigue analysis of the vent line bellows demonstrates their adequacy to accommodate thermal and internal pressure load cycles for the life of the plant. As such this analysis is a TLAA.

Analysis The suppression chamber is in the general form of a torus, which is below and encircles the drywell. The suppression chamber is connected to the drywell by eight vent lines which are connected to a common header. A vent line bellows assembly connects each vent line to the suppression chamber allowing for differential movement between the drywell and the suppression chamber.

Vent line bellows stresses are due primarily to differential thermal expansion of the reactor pressure vessel and the drywell during normal startup and shutdown evolutions and, due to accident conditions. The original vent line bellows was designed and analyzed in accordance with ASME Section III, 1965 Edition including the Summer 1966 Addenda. The current evaluation was performed in accordance with ASME Section III, Subsection NC, using the 1995 Edition including the 1996 Addenda. The current analysis for the vent line bellows conservatively used as the basis for the expected number of cycles to be experienced during the forty (40) year design life 300 startup/shutdown cycles and 1 cycle due to postulated accident conditions.

The result of this analysis was confirmation that cumulative usage factor (CUF) is significantly below the acceptance criteria of 1.0 for the 40 year design life.

Disposition: Validation, 10 CFR 54.21(c)(1)(i)

By inspection of the current analysis, which predicts a maximum 40 year design CUF of 0.10, the fatigue adequacy of the vent line bellows at MNGP is validated. The capacity of the vent line bellows is adequate for the number of transient cycles expected during the extended 60 year operating period.

Umax,40 = 0.10, x 60/40 = Umax,60 = 0.15 < 1.0 Primary Containment Process Penetration Bellows Fatigue Analysis Summary Description Containment pipe penetrations that are required to accommodate thermal movement have expansion bellows. The bellows are designed for a minimum number of operating cycles over the design life of the plant.

Consequently, the primary containment process penetrations bellows cycle basis is a TLAA.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 68 of 84 Analysis At MNGP, the only containment process piping that is subject to significant thermal expansion and contraction are those that penetrate the drywell shell.

Typically these penetrations, which were designed to the ASME Code,Section III, Class B requirements, are a triple flued head design which has a guard pipe between the process piping and the penetration nozzle. This permits the penetration to be vented to the drywell should a rupture of the hot line occur within the penetration.

These containment penetration process bellows have been designed for a minimum of 7,000 operating cycles.

Disposition: Validation, 10 CFR 54.21(c)(1)(i)

Transient cycles on the bellows are composed primarily of thermal cycles experienced by the attached piping. The cycle requirements can be conservatively approximated by the maximum number of thermal cycles specified for any reactor pressure vessel nozzle. For MNGP the limiting nozzle from a total cycle standpoint is the feedwater nozzle, which has as its design basis 1,500 applied cycles for a 40 year operating period. For the 60 year extended operating period, the number of cycles can be estimated by multiplying the 40 year value times 1.5 which results in an estimated design cycle expectation of less than 2,250 or less than one-third of the original design requirement.

Consequently, the current containment penetration bellows fatigue design criteria remain valid with significant margin for the 60 year extended operating period.

K3.9 Environmental Qualification of Electrical Equipment (EQ)

Summary Description 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, specifically requires that an environmental qualification program be established to demonstrate that certain electrical components located in harsh plant environments are qualified to perform their safety function in those harsh environments after the effects of in-service aging.

The MNGP Environmental Qualification Program meets the requirements of 10 CFR 50.49 for the applicable components important to safety.

10 CFR 50.49(e)(5) contains provisions for aging that include consideration of all significant types of aging degradation that can affect component functional capability.

10 CFR 50.49(e) also requires replacement or refurbishment of components qualified for less than the current license term prior to the end of designated life unless additional life is established through ongoing qualification.

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Guidelines (Reference 16), NUREG-0588, Regulatory Guide 1.89, and in Generic Letter 82-09.

The MNGP EQ Program manages component thermal, radiation and cyclical aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. Aging evaluations for EQ components that specify a qualification of at least 40 years are TLAAs for license renewal. The EQ Program will manage the aging effects of applicable components in the EQ program. Section 4.4.2.1.3 of NUREG-1800 states that the staff has evaluated the EQ Program (10 CFR 50.49) and determined that it is an acceptable aging management program to address EQ according to 10 CFR 54.21(c)(1)(iii), Aging Management. This evaluation is documented in NUREG-1801, Generic Aging Lessons Learned (GALL) Report,Section X.E1, Environmental Qualification of Electric Components.

The MNGP EQ Program is an existing program, established to meet commitments for 10 CFR 50.49, that are consistent with NUREG-1801, Generic Aging Lessons Learned (GALL) Report,Section X.E1, Environmental Qualification of Electric Components. In accordance with 10 CFR 54.21(c)(1)(iii),

the EQ Program, which implements the requirements of 10 CFR 50.49, is viewed as an aging management program for license renewal. Reanalysis of an aging evaluation to extend the qualification of components under 10 CFR 50.49(e) is performed as part of the EQ Program at MNGP.

Analysis Aging evaluations of electrical components in the EQ program at MNGP that specify a qualified life of at least forty years are TLAAs.

Aging evaluations are normally performed to extend the qualification by reducing excess conservatism incorporated in the prior evaluation or by including new aging data. While a component life limiting condition may be due to thermal, radiation, or cyclical aging, the majority of component aging limits are based on thermal conditions. Conservatism may exist in aging evaluation parameters such as the assumed ambient temperature of the component, the activation energy, or in the application of a component (e.g. de-energized vs. energized). Important attributes of a reanalysis include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria and corrective actions (if acceptance criteria are not met). These attributes are discussed in more detail below.

• Analytical Methods - The MNGP EQ Program generally uses the same analytical models in the reanalysis of an aging evaluation as those previously applied for the current evaluation. The Arrhenius methodology is an acceptable model for performing a thermal aging evaluation. The analytical method used for a radiation aging evaluation is to demonstrate qualification for the total integrated dose (that is, normal radiation dose for the projected installed life plus accident radiation dose).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 70 of 84 For license renewal, acceptable methods for establishing the 60 year normal radiation dose includes multiplying the 40 year normal radiation dose by 1.5 (that is, 60 years/40 years) or using the actual calculated value for 60 years.

The result is added to the accident radiation dose to obtain the total integrated dose for the component. In some cases, the normal radiation dose is insignificant when compared to the accident dose. In such cases the accident dose may be valid for both the 40 year and 60 year dose. For cyclical aging a similar approach may be used. Other models may be justified on a case-by-case basis.

• Data Collection and Reduction Methods - Reducing excess conservatism in the component service conditions (for example, temperature, radiation, cycles) used in the prior aging evaluation is the primary method used for a reanalysis per the EQ Program. Temperature data used in an aging evaluation should be conservative and based on plant design temperature or on actual plant temperature data. When used, plant temperature data can be obtained in several ways including monitors used for technical specification compliance, other installed monitors, measurements made by plant operators during rounds, and temperature sensors on large motors (while the motor is not running). A representative number of temperature measurements are conservatively evaluated to establish the temperature used in an aging evaluation. Plant temperature data may be used in an aging evaluation in different ways, such as (a) directly applying the plant temperature data in the evaluation or (b) using the plant temperature data to demonstrate conservatism when using plant design temperatures for an evaluation. Any changes to the material activation energy values as part of a reanalysis are to be justified on a plant specific basis. Similar methods of reducing excess conservatism in the component service conditions used in prior aging evaluations can be used for radiation and cyclical aging.

• Underlying Assumptions - EQ component aging evaluations contain sufficient conservatism to account for most environmental changes occurring due to plant modifications and events. When unexpected adverse conditions are identified during operational or maintenance activities that affect the normal operating environment of a qualified component, the affected EQ component is evaluated and appropriate corrective actions are taken, which may include changes to the qualification bases and conclusions.

• Acceptance Criteria and Corrective Action - The reanalysis of an aging evaluation could extend the qualification of the component. If the qualification cannot be extended by reanalysis, the component is maintained, replaced, or re-qualified prior to exceeding the period for which the current qualification remains valid. A reanalysis is performed in a timely manner (that is, sufficient time is available to maintain, replace, or re-qualify the component if the reanalysis is unsuccessful).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 71 of 84 Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)

Based on a review of the MNGP EQ Program and operating experience, the continued effective implementation of the program provides reasonable assurance that (a) the aging effects will be managed, and (b) EQ components will continue to perform their intended function(s) consistent with the current licensing basis for the period of extended operation. Therefore, the MNGP EQ Program is an acceptable aging management program for license renewal under 10 CFR 54.21(c)(1)(iii) during the period of extended operation.

K3.10 Reactor Building Crane Load Cycles Summary Description The MNGP Reactor Building Crane System consists of an 105 ton bridge crane.

The crane is capable of handling the drywell head, reactor vessel head, pool plugs and spent fuel cask. A refueling service platform, with necessary handling and grappling fixtures, services the refueling area and the spent fuel pool.

The Reactor Building Crane System has been modified to incorporate redundant safety features which were not a part of the original design. The modification consists of a new trolley with redundant design features and a capacity of 105 tons on the main hook with redundancy features and a conventional 5 ton capacity hook. This modification was implemented for handling heavy loads both during refueling operations and during operations involving the transfer of spent fuel offsite or to the sites Independent Spent Fuel Storage Installation (ISFSI).

Such spent fuel transfers can take place either when the plant is operating or shut down. The redundant crane has been installed to reduce the probability of a heavy load drop to the category of an incredible event.

NUREG-0612 suggests that cranes should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976, Overhead and Gantry Cranes, and of CMAA-70, Specifications for Electric Overhead Traveling Cranes.

The Reactor Building Crane, manufactured prior to the issuance of CMAA-70 and ANSI B30.2, was designed to meet EOCI 61.

Since the evaluation used as a basis, an expected number of load cycles over the 40 year life of the plant Reactor Building Crane load cycles are a TLAA.

Analysis Reactor Building Crane System design conservatively considers that the following heavy load cycles will be required during the 40 year plant life: 20 lifts per year of Reactor Building shield blocks and plugs, 2 lifts per year of the reactor vessel head, 2 lifts per year of the drywell vessel head, 2 lifts per year of the steam separator assembly and, 2 lifts per year of the steam dryer assembly.

Without consideration for the fact that the modified Reactor Building Crane System was installed after several years of operation the total amount of heavy lifts expected during a 40 year life is 1,120 cycles.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 72 of 84 Disposition: Validation, 10 CFR54.21(c)(1)(i)

The Reactor Building Crane is conservatively designed to handle more than 100,000 heavy loads over the 60 year operating life of the plant. By inspection, the crane is expected to be subjected to less than 2,000 heavy lifts during the 60 year extended operating period, which is significantly less than the design value. Therefore, fatigue life is not significant for the operation of the Reactor Building Crane System and the current analysis remains valid for the period of extended operation.

K3.11 Fatigue Analyses of HPCI & RCIC Turbine Exhaust Penetrations Summary Description To evaluate the effects of testing the operability and performance of the turbine-pump units on a periodic basis MNGP conducted a detailed evaluation of the thermal cycles experienced during testing. Since the number of cycles used in the evaluation is based on a 40 year plant life, this is a TLAA.

Analysis The existing evaluation of the High Pressure Coolant Injection turbine exhaust nozzle used test conditions of 292°F and 50 psig in conjunction with Mark I loads to calculate a cumulative fatigue usage factor. The main conclusion of this evaluation was that the maximum number of High Pressure Coolant Injection turbine tests allowed was only 260, or approximately one test every other month assuming a 40 year plant life.

The major factor was the design temperature of 292°F, the saturated steam temperature associated with the torus at a design pressure of 50 psig. Since the normal operating pressure of the torus is close to atmospheric, it was believed that the actual test temperature was closer to 212°F. To confirm this, the High Pressure Coolant Injection and Reactor Core Isolation Cooling torus nozzles were instrumented to obtain the actual temperature responses during operational testing. The conclusion of these tests was that the maximum temperature that either of these nozzles will experience is expected not to exceed 225°F. A thermal stress analysis was subsequently completed for both nozzles. Finite element models were developed for both nozzles which included explicit modeling of the nozzle to insert plate welds and nozzle to sleeve welds. The evaluation was performed for the following thermal load cases:

• A through wall temperature of the nozzle wall at 225°F with the torus insert plate at 70°F. This corresponds to the initial heatup of the nozzle that occurs immediately after turbine start.

• A through wall temperature of 118°F to simulate a rapid cooldown which occurs during reflood. This corresponds to the average temperature of the Reactor Core Isolation Cooling nozzle immediately after turbine shutdown.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 73 of 84 These two cases were separately evaluated for each penetration. Based on the results, usage factors were calculated in accordance with Section III, Subsection NE of the ASME code. The maximum peak stress ranges for the heatup and cooldown cycles are 77.4 ksi and 83.5 ksi for the High Pressure Coolant Injection and Reactor Core Isolation Cooling penetrations, respectively. Based on an assumption that 676 single safety relief valve (SRV) actuations and 258 multiple valve actuations will occur during the 40 year plant life, the SRV usage factors for the High Pressure Coolant Injection and Reactor Core Isolation Cooling nozzles are 0.009 and 0.043, respectively. The worst case fatigue loading for both nozzles that could be caused by Mark I LOCA loads is a DBA CO acting simultaneously with OBE. One turbine actuation cycle was also postulated for this case. From the Mark I program stress results, the maximum LOCA usage factors for the High Pressure Coolant Injection and Reactor Core Isolation Cooling nozzles are 0.044 and 0.228, respectively. By summing the usage factors for the SRV actuations and Mark I LOCA loads plus OBE, cumulative usage factors of 0.053 and 0.271 were obtained for the High Pressure Coolant Injection and Reactor Core Isolation Cooling nozzles, respectively.

Disposition: Validation, 10 CFR 54.21(c)(1)(i)

Considering that the effects of power rerate implemented in 1998 may increase the design cycles for SRV actuations by as much as 26 percent due to higher steaming rates (conservative percent increase corresponding to 1880 MWt), the maximum contribution due to SRV cycles is 1.26 times 0.043, or 0.054.

Consequently, the maximum cumulative fatigue usage for 40 years is 0.282. This updated, current analysis, therefore, is validated for 60 years of operation by:

Umax,40 = 0.282, x 60/40 = Umax,60 = 0.423 < 1.0 This results in a minimum of 0.577 available fatigue usage due to operational testing of the Reactor Core Isolation Cooling turbine which corresponds to 3,826 operational tests (an average of more than 5 tests per month over the 60 year extended life).

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 74 of 84 K4 TLAA Supporting Activities K4.1 Environmental Qualification (EQ) of Electrical Components The purpose of the MNGP EQ Program is to ensure that safety-related electrical equipment is capable of performing its function in a harsh environment (effects of a loss of coolant accident (LOCA), high energy line break (HELB), or post LOCA radiation) and is qualified in accordance with the Equipment Qualification Final Rule, 10 CFR 50.49, dated February 22, 1983. The MNGP program will continue through the end of the 20-year period of extended operation.

K4.2 Metal Fatigue of Reactor Coolant Pressure Boundary The MNGP Metal Fatigue of the Reactor Coolant Pressure Boundary aging management program is part of the MNGP Thermal Fatigue Monitoring Program.

The MNGP Thermal Fatigue Monitoring Program provides for the periodic review of plant transients for impact on selected components. In addition, MNGP has evaluated environmental effects in accordance with NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves for Selected Nuclear Power Plant Components. Selected components were evaluated using material specific guidance presented in NUREG/CR-6583 for carbon and low alloy steels and in NUREG/CR-5704 for austenitic stainless steels. The MNGP program ensures that limiting components remain within the acceptance criteria for cumulative fatigue usage throughout the licensed term and, that if trends indicate otherwise, appropriate corrective action can be implemented. NSPM will also incorporate requirements for inclusion of NUREG/CR-6260 locations in implementing procedures for the MNGP Thermal Fatigue Monitoring Program.

K4.3 Exemptions The requirements of 10 CFR 54.21(c) stipulate that the application for a renewed license should include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and that are based on time-limited aging analyses, as defined in 10 CFR 54.3. Each active 10 CFR 50.12 exemption has been reviewed to determine whether the exemption is based on a time-limited aging analysis. No existing TLAA related exemptions were identified.

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1. Each year, following the submittal of the MNGP License LRA Section Annually Renewal Application and at least three months before the 1.4 scheduled completion of the NRC review, NSPM will submit amendments to the MNGP application pursuant to 10 CFR 54.21 (b). These revisions will identify any changes to the current licensing basis that materially affect the contents of the License Renewal Application, including the USAR supplements and any other aspects of the application.

2. In accordance with the guidance of Appendix A.3.2.1.2 of LRA Section Annually NUREG-1800, Appendix B of the latest issued supplement to 3.0.7 NUREG-0933 will be reviewed for new GSIs designated as USI-, HIGH-, or MEDIUM- priority. Any identified that involve TLAAs or aging effects for structures and components subject to an aging management review will be included in the annual update of the LRA.

Inspection of the steam dryer is to be accomplished using the LRA Sections As Required

3. guidance provided in BWRVIP-139, Steam Dryer Inspection 2.1.4.2.2, and Raw Evaluation Guidelines, (April 2005), for the MNGP A2.1.12, steam dryer inspections. In the event a new steam dryer is B2.1.12 and installed, NSPM will reevaluate the inspection requirements. Table 3.1.2-3, Note 136

4. The interior of the Diesel Fire Pump House masonry block LRA Table As Required walls is covered with insulation. The Structures Monitoring 3.5.2-8, Note Program will require that the interior surfaces of the walls will 516 be examined if exterior wall surfaces show evidence of significant aging effects.

5. The procedures and training used to limit reactor pressure LRA Section As Required vessel cold over-pressure events will be the same as those 4.2.6 approved by the NRC when MNGP requested approval of the BWRVIP-05 technical alternative for the term of the current operating license. A request for extension for the 60-year extended operating period will be submitted to the NRC prior to the period of extended operation.

6. MNGP site-specific administrative work instructions will be LRA Section Prior to Period applicable to both safety and non-safety related systems, B1.3 of Extended structures and components that are subject to an aging Operation management review consistent with the current licensing basis during the period of extended operation.

7. Site documents that implement aging management activities LRA Section Prior to Period for license renewal will be enhanced to ensure that an AR is B1.3 of Extended prepared in accordance with plant procedures whenever non- Operation conforming conditions are found (i.e., the acceptance criteria is not met).

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8. Revisions will be made to procedures and instructions that LRA Section Prior to Period implement or administer aging management programs and/or of Extended B1.3 activities for the purpose of managing the associated aging Operation effects for the duration of extended operation.

9. The MNGP ASME Section Xl, Subsection IWF Program will be LRA Section Prior to Period enhanced to provide inspections of Class MC components of Extended B2.1.3 supports consistent with NUREG-1801, Chapter III, Section Operation B1.3.

10. The guidance for performing visual bolting inspections LRA Section Prior to Period contained in EPRI TR-104213, Bolted Joint Maintenance & B2.1.4 of Extended Application Guide, and the Good Bolting Practices Handbook Operation (EPRI NP-5067 Volumes I and 2) will be included in the Bus Duct Inspection Program, Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program, Structures Monitoring Program and the System Condition Monitoring Program.

11. The Buried Piping and Tanks Inspection Program will update LRA Section Prior to Period the implementing procedures to include inspections of buried B2.1.5 of Extended components when they are uncovered. Operation

12. The Diesel Fuel Oil Storage Tank, T-44, Internal inspection will LRA Section Prior to Period be added to the list of scheduled Inspections in the Buried B2.1.5 of Extended Piping and Tanks Inspection Program. Operation

13. The Buried Piping and Tanks Inspection Program will be LRA Section Prior to Period revised to include a provision that if evaluations of pipe wall B2.1.5 of Extended thickness show a susceptibility to corrosion, further evaluation Operation as to the extent of susceptibility will be performed.

14. The Buried Piping and Tanks Inspection Program will be LRA Section Prior to Period revised to specify a 10-year buried pipe Inspection frequency. B2.1.5 of Extended Operation

15. The Buried Piping and Tanks Inspection Program will be LRA Section Prior to Period revised to specify a 10-year inspection frequency for Diesel B2.1.5 of Extended Fuel Oil Storage Tank T-44. Operation

16. The Buried Piping and Tanks Inspection Program will be LRA Section Prior to Period revised to include a review of previous buried piping issues to B2.1.5 of Extended determine possible susceptible locations. Operation

17. The Bus Duct Inspection Program will be implemented LRA Section Prior to Period consistent with the appropriate ten elements described in B2.1.6 of Extended Appendix A of NUREG-1 800, Standard Review Plan for Operation Review of License Renewal Applications for Nuclear Power Plants.

Commitment deleted; incorporated and implemented in Prior to Period

18. LRA Section accordance with ASME Section XI, Appendix VIII as mandated of Extended B2.1.8 and modified by 10 CFR 50.55a.

Operation

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 77 of 84 ITEM COMMITMENT SOURCE SCHEDULE The BWR Feedwater Nozzle Program will be enhanced so the LRA Section Prior to Period 19.

regions being inspected are consistent with the B2.1.8 of Extended recommendations of GE NE-523-A71-0594-A, Revision 1. Operation Commitment deleted; incorporated and implemented in LRA Section Prior to Period 20.

accordance with ASME Section XI, Appendix VIII as mandated B2.1.8 of Extended and modified by 10 CFR 50.55a. Operation The repair/replacement guidelines in BWRVIP-16, 19, 44, 45, LRA Section Prior to Period 21.

50, 51, 52, 57, and 58 will be added, as applicable, to the B2.1.12 of Extended MNGP BWR Vessel Internals Program. Operation NSPM will perform top guide grid Inspections using the EVT-1 LRA Section During the 22.

method of examination, for the high fluence locations (grid B2.1.12 Period of beam and beam-to-beam crevice slot locations with fluence Extended exceeding 5.0 x 102° n/cm2). Ten percent (10%) of the total Operation population will be inspected within 12 years with a minimum of 5% inspected within the first 6 years.

A one time inspection will be performed to monitor the effects LRA Section Prior to Period 23.

of corrosion on select portions of closed-cycle cooling water B2.1.13 of Extended systems that perform a pressure-integrity intended function. Operation The Compressed Air Monitoring Program procedures will be LRA Section Prior to Period 24.

revised to include corrective action requirements if the B2.1.14 of Extended acceptance limits for water vapor, oil content, or particulate are Operation not met. Also, the acceptance criteria for oil content testing will be clarified and the basis for the acceptance limits for the water vapor, oil content, and particulate tests will be provided.

The Compressed Air Monitoring Program will be revised to LRA Section Prior to Period 25.

include inspection of air distribution piping based on the B2.1.14 of Extended recommendations of EPRI TR-108147. Operation

26. The MNGP Electrical Cables & Connections Not Subject To LRA Section Prior to Period 10 CFR 50.49 Environmental Qualification Requirements B2.1.15 of Extended Program will be implemented as a new program consistent with Operation the recommendations of NUREG-1801 Chapter Xl Program XI.E1. The program will manage the aging of conductor insulation material on cables, connectors, and other electrical insulation materials that are installed in an adverse localized environment caused by heat, radiation, or moisture.

27. The Electrical Cables Not Subject to 10 CFR 50.49 LRA Section Prior to Period Environmental Qualification Requirements Used in B2.1.16 of Extended Instrumentation Circuits Program will be Implemented as a new Operation program. With exceptions, it will be consistent with the recommendations of NUREG-1 801 Chapter Xl Program XI.E2.

The MNGP Fire Protection Program will be revised to include a LRA Section Prior to Period 28.

visual inspection of the halon fire suppression system to detect B2.1.17 of Extended any signs of degradation, such as corrosion and mechanical Operation damage. This visual Inspection will provide aging management for external surfaces of the halon fire suppression system.

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29. The Fire Protection Program plan document will be revised to LRA Section Prior to Period include qualification criteria for Individuals performing visual B2.1.17 of Extended inspections of penetration seals, fire barriers, and fire doors. Operation The qualification criteria will be based on physical capability (i.e., eye exam), education/training and experience with Fire Protection Program requirements.

30. The Fire Water System Program Implementing procedures will LRA Section Prior to Period be revised to include the extrapolation of inspection results to B2.1.18 of Extended below grade fire water piping with similar conditions that exist Operation within the above grade fire water piping.

The MNGP Fire Water System Program sprinkler heads will be Prior to Period

31. LRA Section inspected and tested per NFPA requirements. Per the NFPA of Extended B2.1.18 code, the sprinkler heads will be tested or replaced when the Operation sprinklers have been in service for 50 years. Testing procedures shall be repeated at 10-year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner. If the sprinkler heads are replaced testing is not required.

32. The Fire Water System Program will verify the procedures to LRA Section Prior to Period be used for aging management activities of the Fire Water B2.1.18 of Extended System apply testing in accordance with applicable NFPA Operation codes and standards. Revise the relevant procedures as appropriate.

33. The MNGP procedures related to the Diesel Fuel Oil System LRA Section Prior to Period will be revised to include requirements to check for general, B2.1.20 of Extended pitting, crevice, galvanic, microbiological influenced corrosion Operation (MIC), and cracking.

34. The MNGP Fuel Oil Chemistry Program procedures will be LRA Section Prior to Period revised to require tank draining, cleaning, and inspection if B2.1.20 of Extended deemed necessary based on the trends Indicated by the results Operation of the diesel fuel oil analysis, or as recommended by the system engineer based on equipment operating experience.

35. Develop or revise existing procedures in the MNGP Fuel Oil LRA Section Prior to Period Chemistry Program to require periodic tank inspections of the B2.1.20 of Extended diesel fuel oil tanks. Operation

36. The MNGP Inaccessible Medium Voltage (2 kV to 34.5 kV) LRA Section Prior to Period Cables Not Subject to 10 CFR 50.49 EQ Requirements B2.1.21 of Extended Program will be implemented as a new program consistent with Operation the recommendations of NUREG-1801 Chapter Xl Program XI.E3.

37. The Inspection of Overhead Heavy Load & Light Load (Related LRA Section Prior to Period to Refueling) Handling Systems Program will be enhanced to B2.1.22 of Extended specify a five-year Inspection frequency for the fuel preparation Operation machines.

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38. The MNGP One-Time Inspection Program will be implemented LRA Section Prior to Period as a new program consistent with the recommendations of B2.1.23 of Extended NUREG-1 801 Chapter Xl Program M32, "One-Time Operation Inspection." This program will include measures to verify the effectiveness of the following aging management programs:

Plant Chemistry Program and Fuel Oil Chemistry Program.

This program will also confirm the absence of age degradation in selected components (e.g., flow restrictors, venturis) within License Renewal scope.

39. The MNGP Protective Coating Maintenance and Monitoring LRA Section Prior to Period Program procedures will be updated to include inspection of all B2.1.27 of Extended accessible painted surfaces Inside containment. Operation

40. The MNGP Protective Coating Maintenance and Monitoring LRA Section Prior to Period Program will be revised to include a pre-inspection review of B2.1.27 of Extended the previous two inspection reports so that trends can be Operation identified.

41. The MNGP Protective Coating Maintenance and Monitoring LRA Section Prior to Period Program implementation procedures will be revised to include B2.1.27 of Extended provisions for analysis of suspected reasons for coating failure. Operation

42. NSPM intends to use the Integrated Surveillance Program for LRA Section Prior to Period MNGP during the period of extended operation by B2.1.29 of Extended implementing the requirements of BWRVIP-86 R1-A, and Operation subsequent NRC Approved Revisions.

43. NSPM will retain the capsules removed from the MNGP reactor LRA Section Prior to and vessel as part of the Reactor Vessel Surveillance Program. B2.1.29 during the Period of Extended Operation

44. The MNGP Selective Leaching of Materials Program will be LRA Section Prior to Period implemented as a new program consistent, with exceptions, to of Extended B2.1.30 the recommendations of NUREG-1 801 Chapter XI Program Operation M33, "Selective Leaching of Materials." The program will be developed and implemented before the start of the period of extended operation. The program includes a one-time visual inspection and hardness measurement of selected components that are susceptible to selective leaching. In situations where hardness testing is not practical, a qualitative method by other NDE or metallurgical methods will be used to determine the presence and extent of selective leaching. The program will determine if selective leaching is occurring for selected components.

45. The MNGP Structures Monitoring Program will be expanded, LRA Section Prior to Period as necessary, to include inspections of structures and structural B2.1.31 of Extended elements in scope for License Renewal that are not inspected Operation as part of another aging management program.

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46. The MNGP Structures Monitoring Program implementing LRA Section Prior to Period procedures will be enhanced to ensure that structural B2.1.31 of Extended inspections are-performed on submerged portions of the intake Operation structure from the service water bays to the wing walls.

47. The MNGP Structures Monitoring Program implementing LRA Section Prior to Period procedures will be revised to include the monitoring/inspection B2.1.31 of Extended parameters for structural components within the scope of Operation License Renewal.

48. The MNGP Structures Monitoring Program will be enhanced to LRA Section Prior to Period include a requirement to sample ground water for pH, chloride B2.1.31 of Extended concentration and sulfate concentration. Operation

49. The MNGP Structures Monitoring Program will be enhanced to LRA Section Prior to Period include concrete evaluations of inaccessible areas if B2.1.31 of Extended degradation of accessible areas Is detected. Operation

50. The MNGP Structures Monitoring Program implementing LRA Section Prior to Period procedures will be enhanced to Include acceptance criteria for B2.1.31 of Extended structural Inspections of submerged portions of the Intake Operation Structure.

51. Implementing instructions and procedures for the System LRA Section Prior to Period Condition Monitoring Program will be revised to describe B2.1.32 of Extended specific age degradation parameters to be monitored and Operation inspected. Acceptance criteria will also be included.

52. Incorporate requirements for inclusion of NUREG/CR-6260 LRA Sections Prior to Period locations in implementing procedures for the MNGP Thermal B3.2 and A4.2 of Extended Fatigue Monitoring Program. Operation

53. The NSPM Fleet Procedure for the Flow Accelerated Corrosion LRA Sections Prior to Period Inspection Program will be revised to include the accepted A2.1.19and of Extended 87.5% of nominal pipe wall thickness for non-safety related B2.1.19 Operation piping as a trigger point for engineering analysis.

54. Prior to the period of extended operation, coating inspectors LRA Sections Prior to Period will meet the requirements of ANSI N45.2.6. A2.1.27and of Extended B2.1.27 Operation

55. NSPM will implement a program at MNGP which is consistent LRA Sections Prior to Period with the NUREG-1801, Rev. 1, XI.E6 "Electrical Cable 3.6, A2.1.34 of Extended Connections Not subject to 10 CFR 50.49 Environmental and B2.1.34 Operation Qualification Requirements."

56. NSPM has inspected the in-core monitoring dry tubes on an LRA Sections During the every-other refueling outage periodicity and will continue to 4.4, A2.1.12 Period of perform this inspection during the period of extended operation, and B2.1.12 Extended per the guidance provided in GE SIL- 409. Operation

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57. The NSPM is an active member of the BWRVIP and will LRA Sections During the continue to follow applicable inspection guidelines and B1.6, A2.1.12 Period of recommendations, which have been reviewed and approved by and B2.1.12 Extended the executive committee of the BWRVIP, throughout the period Operation of extended operation for MNGP.

NSPM will add inspection requirements for the P1, P2, and P3 LRA Sections Prior to Period

58. Core Spray piping welds, at MNGP, in accordance with B1.6, A2.12 of Extended guidance provided in BWRVIP-18, or subsequent revisions. and B2.12 Operation NSPM commits to adhere to BWRVIP inspection guidelines for LRA Sections Prior to Period

59. core plate hold-down bolts by implementation of one or more of 4.8 and A3.6 of Extended the following: Operation

1. Develop an alternative to the inspections identified in the BWRVIP which will, at a minimum, satisfy the intent of the BWRVIP in terms of assuring core plate functional integrity throughout the period of extended operation, NSPM will provide the alternative to the inspection program to the NRC staff for their review and approval at least one year prior to entering the period of extended operations,

2. Inspect the core plate bolts using either UT, some other volumetric inspections, EVT-1 from below the core plate, or other approved inspections in accordance with BWRVIP-25, to assure an adequate number of the core plate bolts are intact to prevent lateral displacement of the core plate, or

3. Install core plate wedges to structurally replace the lateral load resistance provided by the rim hold-down bolts and perform no inspections.

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1. NEDO-32205A, 10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels, Revision 1, February 1994 as transmitted by BWROG-94037, L.A. England (BWROG) to Daniel G. McDonald (NRC), BWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin Analysis - Approved Version, March 21, 1994.

2. 10 CFR 54.3, Requirements for Renewal of Operating Licenses for Nuclear Power Plants- Definitions, May 8, 1995.

3. Letter, S.A. Richard (NRC) to J.F. Klapproth (GE), Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation, (TAC No. MA9891) MFN 01-050, September 14, 2001.

4. C.I. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company), Acceptance For Referencing of EPRI Proprietary Report TR-113596, 'BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74)' and Appendix A, 'Demonstration Of Compliance With The Technical Information Requirements Of The License Renewal Rule (10 CFR 54.21)',

October 18, 2001.

5. EPRI report TR-1003346, Updated BWR Integrated Surveillance Program (ISP)

Implementation Plan, October 2002 (BWRVIP-86-A) (EPRI Proprietary).

6. MNGP Updated Safety Analysis Report, Revision 21, December 15, 2004.

7. Ranganath, S., Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Postulated Loss of Coolant Accident, Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, August 1979, Paper G1/5.

8. Appendix A to Section XI of the ASME Boiler and Pressure Vessel Code, 2004 Edition.

9. EPRI TR-108279, Fracture Toughness and Tensile Properties of Irradiated Austenitic Stainless Steel Components Removed from Service, (BWRVIP-35),

EPRI, Palo Alto CA, June 1997 (EPRI Proprietary Information).

10. Gus C. Lainas (NRC) to Carl Terry (Niagara Mohawk Power Company), BWRVIP Chairman, Final Safety Evaluation of the BWRVIP Vessel and Internals Project BWRVIP-05 Report, (TAC No. M93925), July 28, 1998.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-K Revision 37 APPENDIX K RENEWED OPERATING LICENSE-USAR SUPPLEMENT Page 83 of 84

11. EPRI Report TR-105697, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), September 28, 1995, with supplementing letters of June 24 and October 29, 1996, May 16, June 4, June 13, and December 18, 1997, and January 13, 1998. (EPRI Proprietary Information)

12. Jack R. Strosnider (NRC), to Carl Terry (BWRVIP Chairman), Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, (TAC No. MA3395), March 7, 2000.

13. NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, November 1980

14. NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.

15. EPRI Report 1007824, Integrated Surveillance Program (ISP) Implementation for License Renewal, July 2003 (BWRVIP-116)(EPRI Proprietary).

16. DOR Guidelines, Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors, U.S. Nuclear Regulatory Commission, June 1979.

17. Monticello PUAR, Volume 3, Vent System Analysis, NSP-74-103, Revision 1, November 1982.

18. Monticello PUAR, Volume 5, Safety Relief Valve Discharge Piping Analysis, NSP-74-105, Revision 1, November 1982.

19. Mark I Plant Unique Load Definition, Monticello Nuclear Generating Plant, NEDO-24576, Rev. 1, October 1981

20. Mark I Containment Program Load Definition Report, NEDO-21888, Revision 2 December 1981

21. NRC (P S Tam) letter to NMC (T J OConnor), Issuance of Ammendment to Conform the Renewed Facility Operating License to Reflect Transfer of Operating Authority (TAC No. MD9621), dated September 22, 2008.

22. NUREG-1865, Safety Evaluation Report, Related to the License Renewal of the Monticello Nuclear Generating Plant, October, 2006.

23. Structural Integrity Associates Inc, Letter Report, TJG-10-004, June 2, 2010

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24. Application for Renewed Operating License, Monticello Nuclear Generating Plant, March 2005

25. SIR-05-044-A, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, Rev. 0, dated April 2007

26. NRC letter to NSP (Mark A. Schimmel), Monticello Nuclear Generating Plant -

Issuance of Amendment to Revise and Relocate Pressure Temperature Curves to a Pressure Temperature Limits Report (TAC NO. ME7930), dated February 27, 2013

27. Calculation 11-003, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, Rev. 1

28. Monticello Pressure Temperature Limits Report (PTLR) up to 54 Effective Full Power Years (EFPY)

29. NRC (T A Beltz) letter to NSPM (K D Fili), "Monticello Nuclear Generating Plant -

Issuance of Amendment No. 176 to Renewed Facility Operating License Regarding Extended Power Uprate (TAC NO. MD9990)", dated December 9, 2013