ML12237A321

License Amendment Request: Revise Renewed Facility Operating License & Technical Specifications to Clarify Fuel Storage Capacity, Remove Obsolete Information and Make Minor Corrections and Miscellaneous Editorial Changes

ML12237A321
Person / Time
Site:	Monticello
Issue date:	08/21/2012
From:	Schimmel M A Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-12-055
Download: ML12237A321 (27)

Text

Monticello Nuclear Generating Plant XceIEnergy~

2807 W County Road 75 Monticello, MN 55362 August 21, 2012 L-MT-12-055 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 License Amendment Request: Revise Renewed Facility Operating License and Technical Specifications to Clarify Fuel Storage Capacity, Remove Obsolete Information and Make Minor Corrections and Miscellaneous Editorial Changes In accordance with 10 CFR 50.90, the Northern States Power Company -Minnesota (NSPM), doing business as Xcel Energy, Inc., proposes to revise the Monticello Nuclear Generating Plant (MNGP) Renewed Facility Operating License and Technical Specification (TS) to reflect editorial corrections to the operating license, make changes in the utilized fuel storage capability, make minor corrections, and remove obsolete information.

NSPM requests U. S. Nuclear Regulatory Commission approval of the proposed license amendment request by September 15, 2013. NSPM requests a 90 day implementation period for this license amendment.

Enclosure 1 provides a description of the proposed changes and includes the technical evaluation and associated no significant hazards determination and environmental evaluations.

Enclosure 2 provides a marked-up copy of the Renewed Facility Operating License pages showing the proposed changes. Enclosure 3 provides a marked-up copy of the TS pages showing the proposed changes.The MNGP Plant Operations Review Committee has reviewed this application.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Minnesota Official.Summary of Commitments This letter proposes no new commitments and does not revise any existing commitments.

AVAL Document Control Desk L-MT-12-055 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.Executed on August Z( , 2012.Mark A. Schimmel Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company -Minnesota Enclosures (3)cc: Administrator, Region III, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce L-MT-12-055 Table of Contents Page 1 of 1 TABLE OF CONTENTS SECTION TITLE PAGE 1.0

SUMMARY

DESCRIPTION 1

2.0 BACKGROUND

1 3.0 DETAILED DESCRIPTION 2 4.0 TECHNICAL ANALYSIS 3 4.1 Revise the Renewed Facility Operating License to 3 Reference the Northern States Power Company -Minnesota or NSPM 4.2 Revise License Condition 2.B.2 to Remove Outdated 4 Reference to SFP Capacity Letter 4.3 Correct Incorrect Phrase in the Core Operating Limits 5 Report Specification

4.4 Remove

OPRM Monitoring Period Note in Table 3.3.1.1-1 5 4.5 Removal of Analytical Methods No Longer Utilized from the 6 Core Operating Limits Report Specification

5.0 REGULATORY

ANALYSIS 6 5.1 No Significant Hazards Determination 6 5.2 Applicable Regulatory Requirements 8 6.0 ENVIRONMENTAL EVALUATION 11

7.0 REFERENCES

12 L-MT-12-055 Enclosure 1 Page 1 of 12 Revise Renewed Facility Operating License and Technical Specifications to Clarify Fuel Storage Capacity, Remove Obsolete Information and Make Minor Corrections and Miscellaneous Editorial Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, the Northern States Power Company -Minnesota (NSPM), doing business as Xcel Energy, Inc., proposes to revise the Monticello Nuclear Generating Plant (MNGP) Renewed Facility Operating License (DPR-263) and several Technical Specifications (TS) to reflect editorial and minor corrections, changes to clarify the fuel storage capacity, and remove obsolete information.

1) Correct or eliminate as applicable, several references to the "Nuclear Management Company, LLC" or "NMC" within the Renewed Facility Operating License (OL) by revising to state "Northern States Power Company", or "NSPM" as applicable.

Revise OL cover page, Appendix A (TS) cover page, and Appendix C -Additional Conditions, to indicate the license was "renewed".

2) Revise License Condition 2.B.2 of the Renewed Facility OL to remove the reference to the August 17, 1977, Northern States Power letter that specifies SFP fuel assembly storage capacity.3) Remove reference in Note (e) to the Oscillation Power Range Monitor monitoring period in Table 3.3.1.1-1, "Reactor Protection System Instrumentation," since this monitoring period has expired.4) Correct terminology for the Average Power Range Monitor (APRM) Simulated Thermal Power (STP) -High function, for single loop operation (i.e., Note b), for"Delta W" under the references in the Core Operating Limits Report (COLR)specification (Specification 5.6.3, Item a.5).5) Remove two analytical methods from Specification 5.6.3, "Core Operating Limits Report (COLR)," that are no longer utilized for licensing basis analyses.6) Editorial and typographical corrections in the Renewed Facility OL and Technical Specifications.

2.0 BACKGROUND

In several places in the Renewed Facility Operating License it has been identified that references to the Nuclear Management Company or NMC have been incorrectly retained.

Also, on several pages the title should have been changed to the Renewed Facility Operating License.

L-MT-1 2-055 Enclosure 1 Page 2 of 12 License Amendment No. 34 revised the MNGP Provisional Operating License (POL)to reflect the revised spent fuel pool (SFP) capacity since the MNGP custom TS did not include a specification prescribing the capacity.

This was necessary due to the re-racking of the SFP with high-density spent fuel storage racks that increased the licensed capacity from 740 to 2237 fuel storage locations (Reference 1). The reference to the letter specifying the SFP capacity under License Condition 2.B.2 can be removed since spent fuel storage capacity information is now included within and controlled directly under the MNGP TS, with the adoption of the Improved Standard Technical Specifications (ITS).License Amendment 159 (Reference

2) revised the TS to reflect the revisions

/additions required to several Specification 3.3.1.1, "Reactor Protection System Instrumentation," functions to reflect adoption of the Power Range Neutron Monitoring System at the MNGP. An inconsistency between the Function 5.b, APRM STP -High equation for single loop operation, presented as Note (b) to Table 3.3.1.1-1, and its description in a listing under the COLR specification (Specification 5.6.3, Item a.5)where the adjustment term "Delta W" is incorrectly referred to as an "Allowable Value", has been identified and is proposed to be corrected.

Also, a 90 day monitoring period was specified for the Operating Power Range Monitoring (OPRM) System reactor stability monitoring system added by Amendment 159 during which the trip outputs were not enabled to provide time to review system functioning without the risk of spurious operation.

This monitoring period is complete and hence this note may be removed.Also, Specification

5.6.3 lists

two analytical methods to determine the core operating limits that are no longer applicable and are proposed to be removed.3.0 DETAILED DESCRIPTION The following changes are proposed to the MNGP Renewed Facility Operating License and TS.1) Revise the front page of the Renewed Facility OL to remove the reference to the"Nuclear Management Company, LLC", and add "RENEWED" in front of"FACILITY" (see Section 4.1).2) Revise License Condition 2.B.2 of the Renewed Facility OL to remove the reference to the August 17, 1977, Northern States Power letter that specifies SFP fuel assembly storage capacity (see Section 4.2).3) Correct typographical error in License Condition 2.B.5 of the Renewed Facility OL by changing "hall" to "shall" immediately after NSPM in the first sentence.

L-MT-1 2-055 Enclosure 1 Page 3 of 12 4) Revise the cover page to Appendix A, the MNGP Technical Specifications, to remove the reference to the "Nuclear Management Company, LLC" and add"RENEWED" in front of "FACILITY" (see Section 4.1).5) Revise Appendix C, "Additional Conditions," page C-1, to the Renewed Facility OL, to change the reference from the "Nuclear Management Company, LLC", to"Northern States Power Company Minnesota" and add "RENEWED" in front of"FACILITY" (see Section 4.1).6) Revise Appendix C page C-3, to the Renewed Facility OL, to change the reference from "NMC" to "NSPM". (see Section 4.1).7) Revise the description of Function 2.b, "APRM Simulated Thermal Power -High Note b," listed under TS Specification 5.6.3, Item a.5, to remove the term"Allowable" from the statement, and change "Value" to "value" (see Section 4.3).8) Remove description under Note (e) to TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," pertaining to the OPRM Monitoring Period and replace with the phrase "(Not Used.)" since this note is no longer applicable (see Section 4.4).9) Remove Items b.2 and b.3 under TS Specification 5.6.3, which refer to licensing topical reports NSPNAD-8608-A and NSPNAD-8609-A, respectively, since these analytical methods are no longer applicable and replace with the phrase "(Not Used.)" (see Section 4.5).10) Add "s" to the word "exception" -last word in the Specification 5.5.11 .a paragraph.

The proposed changes to the Renewed Facility Operating License are provided in Enclosure

2. The proposed changes to the MNGP TS (mark-ups) are provided in Enclosure

3. There are no changes to the MNGP TS Bases.4.0 TECHNICAL ANALYSIS The following proposed changes to the MNGP Renewed Facility Operating License and Technical Specifications reflect clarifications and changes to the utilized fuel storage capability, and several minor corrections and editorial changes to the licensing basis.4.1 Revise the Renewed Facility Operating License to Reference the Northern States Power Company -Minnesota or NSPM Several administrative

/ editorial changes to the MNGP Renewed Facility Operating License are included in Enclosure

2. Following license transfer from L-MT-12-055 Enclosure 1 Page 4 of 12 the Nuclear Management Company to NSPM, it was identified that there were several places in the Renewed Facility Operating License where the Nuclear Management Company, LCC title, or NMC acronym, had been inadvertently retained.

Also, when the Facility Operating License was renewed, the title for the license on several pages should have been changed to the Renewed Facility Operating License. To correct this, several administrative

/ editorial changes, as discussed under Section 2.0, Detailed Description, are proposed.

A mark-up of the proposed changes to the Renewed Facility Operating License is included in Enclosure

2.4.2 Revise

License Condition 2.B.2 to Remove Outdated Reference to SFP Capacity Letter The original capacity of the MNGP SFP was 740 fuel assemblies.

License Amendment No. 34 to the MNGP Provisional Operating License (POL) approved re-racking of the SFP with high-density spent fuel storage rack modules increasing the capacity of the SFP from 740 to 2237 fuel storage locations (Reference 1). The TS in effect at the time, the custom TS, pre-date the current MNGP TS (in the ITS format) and the format did not specify the licensed capacity of the SFP. Accordingly, License Condition 2.B.2 of the POL was amended to reflect NRC approval by referencing an August 17, 1977, Northern States Power letter that in-part specified the SFP fuel assembly storage capacity.

License Condition 2.B.2 contains a reference to this letter that is outdated and can be removed. The phrase struck-out below is proposed to be removed. License Condition 2.B.2 states: Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 1974 (those portions dealing with handling of reactor fuel) and August-!7, 1977 (those portions dealing with fuel assembly storage capacity);

Since this August 17, 1977 letter, a license amendment request (LAR) to increase SFP storage capacity have been submitted and approved by the NRC.This LAR and conversion to the Improved Standard Technical Specifications (ITS) have resulted in changes to TS Section 4.0, "Design Features," related to fuel storage. License Condition 2.B.2 is proposed to be revised to reflect these changes (see strikethrough phrase to be removed above) and NSPM proposes to remove this outdated reference.

A mark-up with this proposed change to the Renewed Facility Operating License is included in Enclosure

2.

L-MT-12-055 Enclosure 1 Page 5 of 12 4.3 Correction of Incorrect Phrase in the Core Operating Limits Report Specification Specification 5.6.3, "Core Operating Limits Report (COLR)," describes the specifications and functions for which operating limits are prescribed and indicates the licensing topical reports providing the analytical limits for their determination.

License Amendment 159 (Reference 2), revised the COLR to reflect new methodologies adopted with the approval of the application of the NUMAC Power Range Neutron Monitoring System (PRNMS) and added/revised several TS instrumentation functions.

As part of this change, Note (b) to TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," was revised to reflect a new equation for the APRM Simulated Thermal Power -High function for single loop operation where Delta W is specified in the COLR. Note (b) was revised to state: (b) < 0.66 (W -Delta W) + 61.6% RTP, when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." The cycle-specific value for Delta W is specified in the COLR.Note (b) to Table 3.3.1.1-1 above, correctly indicates that the cycle-specific value for Delta W is specified in the COLR. However, the title for this function as specified in the COLR specification, below, is incorrect because it implies that Delta W is an allowable value. Specification 5.6.3, Item a.5 states: Reactor Protection System Instrumentation Delta W Allowable Value for Table 3.3.1.1-1, Function 2.b, APRM Simulated Thermal Power- High, Note b; and Consequently, NSPM proposes to revise the description of Function 2.b, "APRM Simulated Thermal Power -High Note b," listed under Specification 5.6.3, Item a.5, to remove "Allowable" from the statement and change "Value" to"value". This correction makes the terminology in the Reactor Protection System Instrumentation table (Table 3.3.1.1-1) and Specification 5.6.3 consistent.

A mark-up with this proposed change to the TS is provided in Enclosure

3.4.4 Remove

OPRM Monitoring Period Note in Table 3.3.1.1-1 As part of the implementation of the PRNMS, a note was added to Table 3.3.1.1-1, "Reactor Protection System Instrumentation," to specify and Operating Power Range Monitoring (OPRM) System monitoring period, since this was a reactor stability monitoring system new to the MNGP. A 90 day OPRM Monitoring Period was requested in the PRNMS license amendment application, during which the system would not be enabled to provide a trip, to gain confidence with the performance of the system and prevent spurious trips.Note (e) to Table 3.3.1.1-1, states:

L-MT-1 2-055 Enclosure 1 Page 6 of 12 (e) During the OPRM Monitoring Period the OPRM Upscale function is inoperable.

The OPRM Monitoring Period for implementation of the PRNMS is complete and the system is in full operation.

This note is no longer applicable and should be removed from the TS. A mark-up with this proposed change to the TS is provided in Enclosure

3.4.5 Removal

of Analytical Methods No Longer Utilized from the Core Operating Limits Report Specification Item b, under Specification 5.6.3, "Core Operating Limits Report (COLR)," lists the reports providing the analytical methods to determine the core operating limits. Two analytical methods, no longer utilized for determination of the licensed operating limits, are proposed to be removed. They are:* NSPNAD-8608-A, "Reload Safety Evaluation Methods for Application to the Monticello Nuclear Generating Plant"." NSPNAD-8609-A, "Qualification of Reactor Physics Methods for Application to Monticello".

This proposed change revises the plant TS to reflect the present analytical licensing basis for determining core operating limits.5.0 REGULATORY ANALYSIS 5.1 No Significant Hazards Determination In accordance with the requirements of 10 CFR 50.90, the Northern States Power Company -Minnesota (NSPM) requests an amendment to facility Renewed Operating License DPR-22, to revise the Monticello Nuclear Generating Plant (MNGP) Renewed Facility Operating License and the Technical Specifications (TSs) to clarify fuel storage capacity, remove obsolete information and make minor corrections and miscellaneous editorial changes.The NSPM has evaluated the proposed change to the TS in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined that operation of the MNGP in accordance with the proposed amendment presents no significant hazards. NSPM's evaluation against each of the criteria in 10 CFR 50.92 follows.

L-MT-1 2-055 Enclosure 1 Page 7 of 12 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.The MNGP TS and Updated Safety Analysis Report (USAR) provide the specific limitations on the number of fuel assemblies in the MNGP spent fuel pool, fresh fuel storage vault, and the reactor core. Removing the outdated letter reference from License Condition 2.B.2 in the Renewed Operating License (ROL) has no effect on these limitations or on the supporting evaluations.

The proposed changes to the TS and ROL are administrative or editorial in nature and do not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected.

The proposed changes do not impact the initiators or assumptions of analyzed events, nor do they impact mitigation of accidents or transient events.Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No The MNGP TS and USAR provide the specific limitations on the number of fuel assemblies in the MNGP spent fuel pool, fresh fuel storage vault, and the reactor core. Removing the outdated letter reference from the license condition in the ROL has no effect on these limitations or on the supporting evaluations.

This proposed change does not introduce a new mode of plant operation and does not involve a physical modification to the plant. The change will not introduce new accident initiators or impact the assumptions made in a safety analysis.The proposed changes to the TS and ROL are administrative in nature and do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, or impact the function of plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected.

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

L-MT-12-055 Enclosure 1 Page 8 of 12 3. Does the proposed change involve a significant reduction in a margin of safety?Response:

No Margin of safety is related to confidence in the ability of the fission product barriers to perform their design functions during and following postulated accidents.

The MNGP TS and USAR provide the specific limitations on the number of fuel assemblies in the spent fuel pool, fresh fuel storage vault, and the reactor core. Removing the outdated letter reference from the license condition in the ROL has no effect on these limitations or on the supporting evaluations.

Accordingly, no margin of safety is affected.The proposed changes are administrative in nature and do not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected.

The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting conditions for operation, or design parameters for any SSC. The proposed changes do not impact any safety analysis assumptions and do not involve a change in initial conditions, system response times, or other parameters affecting an accident analysis.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based on the above, the NSPM has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92(c), in that it does not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.5.2 Applicable Requlatory Requirements 10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in the TSs. As stated in 10 CFR 50.36, the TSs will include limiting conditions for operation (LCO) (and associated remedial actions)are met.The MNGP was designed largely before the publishing of the 70 General Design Criteria (GDC) for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission for public comment in July 1967, and constructed prior to the 1971 publication of Appendix A, "General Design Criteria for Nuclear L-MT-12-055 Enclosure 1 Page 9 of 12 Power Plants", to 10 CFR Part 50. As such, the MNGP was not licensed to the Appendix A, General Design Criteria (GDC).The MNGP USAR, Section 1.2, lists the principal design criteria (PDCs) for the design, construction and operation of the plant. USAR Appendix E provides a plant comparative evaluation with the proposed AEC 70 design criteria.

It was concluded that the plant conforms to the intent of the GDCs. Therefore, the applicable GDCs are discussed below.PDC 1.2.2.b -- Reactor Core The reactor core, in conjunction with other design parameters, is designed so there is no inherent tendency for sudden divergent oscillation of operating characteristics in any mode of operation.

AEC Draft 1967 GDC 7 -Suppression of Power Oscillations (Category B)The core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

Appendix A, GDC 12 -- Suppression of reactor power oscillations.

The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

PDC 1.2.5.c -- Plant Instrumentation and Control A reliable reactor protection system, independent from the reactor process control system, is provided to automatically initiate appropriate action whenever plant conditions approach pre-established limits. Periodic testing capability is provided.

Sufficient redundancy is provided so that failure or removal from service of any one component or portion of the system will not preclude appropriate actuation of the reactor protection system when required.AEC Draft 1967 GDC 12 -Instrumentation and Control Systems (Category B)Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.Appendix A, GDC 13 -Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

L-MT-12-055 Enclosure 1 Page 10 of 12 AEC Draft 1967 GDC 14 -Core Protection Systems (Category B) Core protection systems together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.Appendix A, GDC 20 -Protection system functions.

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.PDC 1.2.9 -Plant Fuel Handling and Storage Appropriate plant fuel handling and storage facilities are provided to preclude accidental criticality and to provide cooling for the spent fuel.AEC Draft 1967 GDC 66 -Prevention of Fuel Storage Criticality (Category B)Criticality in new and spent storage shall be prevented by physical systems or processes.

Such means as geometrically safe configurations shall be emphasized over procedural controls.Appendix A, GDC 61 -- Fuel storage and handling and radioactivity control.The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

NSPM has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria.

Based on this review, there is reasonable assurance that the health and safety of the public, following approval of this change, is unaffected.

L-MT-12-055 Enclosure 1 Page 11 of 12 6.0 ENVIRONMENTAL EVALUATION The NSPM has determined that the proposed change would not revise a requirement with respect to installation or use of a facility or component located within the restricted area, as defined in 10 CFR 20, nor would it change an inspection or surveillance requirement.

The proposed amendment does not involve (i) a significant hazards consideration, or (ii) authorize a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) result in a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, NSPM concludes that pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

L-MT-12-055 Enclosure 1 Page 12 of 12

7.0 REFERENCES

1. Letter from U.S. NRC to L. 0. Mayer (NSP) issuing Amendment No. 34 to Provisional Operating License No. DPR-22, for the Monticello Nuclear Generating Plant, dated April 14, 1978.2. Letter from U.S. NRC to J. T. Conway (NMC), "Monticello Nuclear Generating Plant (MNGP) -Issuance of Amendment Regarding the Power Range Neutron Monitoring System (TAC No. MD8064)," January 30, 2009.

ENCLOSURE 2 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS TO CLARIFY FUEL STORAGE CAPACITY, REMOVE OBSOLETE INFORMATION AND MAKE MINOR CORRECTIONS AND MISCELLANEOUS EDITORIAL CHANGES MARKED-UP RENEWED FACILITY OPERATING LICENSE PAGES (6 pages follow)

CFACILITY OPERATING LICENSE DPR-22 FOR MONTICELLO NUCLEAR GENERATING PLANT UNIT 1 MONTICELLO, MINNESOTA NORTHERN STATES POWER COMPANY NUCLEARO MANAGEMENT COMPANY. LLC5 DOCKET NO. 50-263 November 8, 2006 2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated Au 16 1974 (those portions dealing with handling of reactor fuel)rand August 17, 1977 (those portions dealing with fuel assembly storage capacity);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or Instrument calibration or associated with radioactive apparatus or components; and 5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: 1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1775 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 167, are hereby incorporated in the license. NSPM shall I operate the facility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans Including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 1-th-u 167 Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p)(2).

The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Monticello Nuclear Generating Plant Physical Security, Training and Qualification, and Safeguards Contingency Plan," with revisions submitted through May 12, 2006.NSPM shall fully implement and maintain in effect all provisions of the Commission-approved.

Northern States Power Company -Minnesota (NSPM) Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The NSPM CSP was approved by License Amendment No. 166.4. Fire Protection NSPM shall implement and maintain in effect all provisions of the'approved fire protection program as described in the Updated Safety Analysis Report for the facility and as approved in the SER dated August 29, 1979, and supplements dated February 12, 1981. and October 2, 1985, subject to the following provision:

NSPM may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.5. Emergency Preparedness Plan NSPM b4 follow and maintain in effect emergency plans which meet the standards of 10 CFR 50.47(b) and the requirements in 10 CFR 50, ,/i t*..I IAppendix E, including amendments and changes made pursuant to the authority of 10 CFR 50.54(q).

The licensee shall meet the requirements of 10 CFR 50.54(s), 50.54(t), and 50.54(u).6. TMI Action Plan NSPM has satisfactorily met all TMI-2 Lessons Learned Category "A" requirements applicable to the facility.

NSPM shall make a timely submittal in response to the letter dated October 31, 1980 regarding post-TMI requirements from Darrell G. Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits (NUREG-0737).

7. Repairs to the Recirculation System Piping The repairs to the recirculation system piping are approved and the unit is hereby authorized to return to power operation, subject to the following condition:

Prior to the startup of Cycle 11, NSPM shall submit by August 1, 1983 for the Commission's review and approval, a program for inspection and/or modification of the recirculation system piping.Renewed License No. DPR-22 Amendment No. 166 APPENDIX A TO fg LA) E WE aD --)4 FACILITY OPERATING LICENSE DPR-22 TECHNICAL SPECIFICATIONS FOR MONTICELLO NUCLEAR GENERATING PLANT UNIT 1 MONTICELLO.

MIINNESOTA NORTHERN STATES POWER COMPANY (NUCLEAR MANAGEMENT COMPANY. LLC)DOCKET NO. 50-263 October 29, 2006 ADDpwniyI' r R E EAEV D ........ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. DPR-22 Nuclear Management Company, LLC hall comply with the following conditions on the Qsc hedules noted Delow: Mor-fhAZP .St~ter f'OIAje COMP&Mft I I Amendment Number Additional Condition 98 The emergency operating procedures (EOPs) shall be changed to require manual Isolation of torus and drywell sprays prior to the point where primary containment pressure would not provide adequate net positive suction head (NPSH) for the emergency core cooling system (ECCS) pumps, change the caution statement regarding NPSH In the Primary Containment Pressure EOP to include the core spray pumps, and add a caution statement regarding NPSH considerations for pressure control while venting to control primary containment pressure.98 Finalize the additional containment analysis and associated NPSH evaluation which extends the existing long-term cause evaluation to the time when the required containment overpressure returns to atmospheric conditions.

Changes to the requested long-term containment overpressure, if any, shall be promptly reported to the NRC prior to starting up the unit from the current maintenance outage.98 Submit the results of the additional containment analysis and associated NPSH evaluation discussed above.Implementation Date Prior to starting up from the current maintenance outage, or August 1, 1997, whichever is later.Prior to starting up from the current maintenance outage, or August 1, 1997, whichever is later.Within 90 days from the date of the plant startup from the current maintenance outage, or November 1, 1997, whichever is later.C-1 Amendment No. a&, 49, APPENDIX C -continued Amendment Number Implementation Date Additional Condition 102 All affected process computer and SPDS data points shall be changed to reflect uprate operating conditions 102 Control room simulator changes shall be completed in accordance with ANSI/ANS 3.5-1985 Section 5.4.1, Simulator Performance Testing, and Monticello simulator configuration control procedures.

102 Classroom and simulator training on new knowledge and abilities associated with the power uprate shall be provided in accordance with Monticello Training Center procedures.

NJS PPI 102 (NM shall monitor plant operational parameters 12 oruprate Impacts on the PRA models 102 Control room simulator changes shall be verified against actual plant startup data.Prior to implementation of Amendment No. 102 (prior to exceeding 1670 MWt).Prior to implementation of Amendment No. 102 (prior to exceeding 1670 MWt).Prior to implementation of Amendment No. 102 (prior to exceeding 1670 MWt).During and after the power uprate ascension test program.Within 3 months of completion of the power uprate ascension test program.Within 6 months of completion of the power uprate ascension test program.Within 9 months of completion of the power uprate ascension test program.I 102 The applicable training programs and the simulator shall be modified, or appropriate compensatory actions shall be taken, In accordance with the Monticello Training Center procedures to reflect issues and discrepancies Identified during startup testing.102 The MNGP USAR shall be updated to reflect the changes associated with power uprate operation.

This update shall not Include credit for suppression pool scrubbing in the MSIV leakage pathway in the revised LOCA analysis.C-3 Amendment No. 402 110 ENCLOSURE3 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS TO CLARIFY FUEL STORAGE CAPACITY, REMOVE OBSOLETE INFORMATION AND MAKE MINOR CORRECTIONS AND MISCELLANEOUS EDITORIAL CHANGES MARKED-UP TECHNICAL SPECIFICATION PAGES (4 pages follow)

RPS Instrumentation 3.3.1.1 Table 3.3:1.1-1 (page 1 of 4)Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCtE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE 1. Intermediate Range Monitors a. Neutron Flux- High 2 b. Inop. 2 3 3 3 3 3cC)5(a)G SR 3.3.1.1.1 SR 3.3.1.1.3 SR 3.3.1.1.4 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.14 H SR 3.3.1.1.1 SR 3.3.1.1.3 SR 3.3.1.1.4 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.14 G SR 3.3.1.1.3 SR 3.3.1.1.4 SR 3.3.1.1.12 H SR 3.3.1.1.3 SR 3.3A1.1.4 SR 3.3.1.1.12 G SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.11 SR 3.3.1.1.15 F SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.11 SR 3.3.1.1.15

< 122/125 divisions of full scale< 122/125 divisions of full scale NA NA20% RTP0.66 W+61.6% RTP(b)and116% RTP 2. Average Power Range Monitors a. Neutron Flux- High, (Setdown)b. Simulated Thermal Power- High 2 1 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.(b) ; 0.66 (W- Delta W) +61.6% RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating," The cycle-specific value for Delta W Is specified in the COLR.(c) Each APRM / OPRM channel provides Inputs to both trip systems.Monticello 3.3.1.1-6 Amendment No. 4466, 159 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 4)Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.I REQUIREMENTS VALUE c. Neutron Flux- High 1 3(c) F SR 3.3.1.1.1

<122% RTP SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.11(0(9)

SR 3.3.1.1.15

d. Inop. 1,2 3(c) G SR 3.3.1.1.4 NA SR 3.3.1.1.15

e. 2-Out-Of-4Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.4 SR 3.3.1.1.12 SR 3.3.1.1.14 2SR 3.3.1.1.15

f. OPRM Upscal[D 20% RTP 3(C) I SR 3.3.1.1.1 As specified SR 3.3.1.1.4 in COLR SR 3.3.1.1.6 SR 3.3.1.1,11 SR 3.3.1.1.15 SR 3.3.1.1.16

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.4 1075 psig Dome Pressure -High .SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.14

4. Reactor Vessel Water 1.2 2 G SR 3.3.1.1.1 inches Level -Low SR 3.3.1.1.4 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.14 (c) Each APRM OPRM channel provides Inputs to both trip systems. 9"-" (e) LDuring the OPRM Monitoring Period the OPRM Upscale function Is inoperable. , .(f) If the as-found channel setpoint is not the Nominal Trip Setpolnt but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it Is functioning as required before returning the channel to service.(g) The Instrument channel setpoint shall be reset to the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, týhe channel shall be declared Inoperable.

The NTSP and the methodology used to determine the NTSP are specified in the Technical Requirements Manual.Monticello 3.3.1.1-7 Amendment No. 446, 159

• Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Safety Function Determination Program (SFDP) (continued)

3. A required system redundant to the support system(s) for the-supported systems described in Specifications 5.5.10.b.1 and 5.5.10.b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.5.5.11 Primary Containment Leakage Rate Testing Program a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"detd September, 1995,as modified by the following exceptio ,j) a 1. The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in March 1993;2. The main steam line pathway leakage contribution is excluded from the sum of the leakage rates from Type B and C tests specified in Section l1l.B of 10 CFR 50, Appendix J, Option B, Section 6.4.4 of ANSI/ANS 56.8-1994, and Section.10.2 of NEI 94-01, Rev. 0; and 3. The main steam line pathway leakage contribution is excluded from the overall integrated leakage rate from Type A tests specified in Section III.A of 10 CF.R 50, Appendix J, Option B, Section 3.2 of ANSI/ANS 56.8-1994, and Section 8.0 and 9.0 of NEI 94-01, Rev. 0.b. The calculated peak containment internal pressure for the design basis loss-of coolant accident, P., is 42 psig. The containment design pressure Is 56 psig.c. The maximum allowable containment leakage rate, La, at Pa, shall be 1.2%of containment air weight per day.Monticello

.5.5-10 Amendment No. 4-46 148 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

4. Control Rod Block Instrumentation Allowable Value for the Table 3.32.1-1 Rod Block Monitor Functions l.a, 1.b, and 1.c and.associated Applicability RTP levels;5. Reactor Protection System Instrumentation Delta W6 alue (for Table 3.3.1.1-1, Function 2.b, APRM Simulated Thermal Power-High, Note b; and 6. Reactor Protection System Instrumentation Period Based Detection Algorithm trip setpoints associated with Table 3.3.1.1-1, Function 2.f, OPRM Upscale.b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-2401 i-P-A, "General Electric Standard Application for Reactor Fu~el";CN U.se.,) 2 1 fNSP-NAD-86,08-A, "Reload Safety Evaluation Methods for Applicatio Dn_to the Monticello Nuclear Generating Plant"; -.k/ eh, \__,) ("NSPNAD-8609-A, "Qualification of Reactor Physics Methods for 6u~i- ~ -J pplication to Monticello";-

4. NEDO-31960, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology";

and 5. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996.The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.Monticello 5.6-2 Amendment No. 4.46~, 4~...Monticello 5.6-2 Amendment No. 446a, 4*9-