Text

Monticello - License Amendment Request: Maximum Extended Load Line Limit Analysis Plus
	ML100280558
Person / Time
Site:	Monticello
Issue date:	01/21/2010
From:	O'Connor T J; Northern States Power Co, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-MT-10-003
	Download: ML100280558 (49)
	v • d • e

Xoel Energyo WITHHOLD Attachment 3 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 January 21, 2010 L-MT-10-003 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 License No. DPR-22 License Amendment Request: Maximum Extended Load Line Limit Analysis Plus

References:

1) Letter from NRC to NSPM, "Monticello Nuclear Generating Plant -Revised Schedule for Review of Extended Power Uprate Amendment Application (TAC No. MD9990)," dated October 1, 2009.2) NSPM Letter to NRC, "Xcel Energy Request for NRC Concurrent Review of Monticello Nuclear Generating Plant Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

License Amendment Request (LAR) with Extended Power Uprate (EPU) LAR Review Delay (L-MT-09-100)," dated October 28, 2009.3) Letter from NRC to NSPM, "Monticello Nuclear Generating Plant -Linking of the Proposed Extended Power Uprate Amendment and the MELLLA+ Amendment (TAC NOS. MD9990 and ME2449)," dated November 23, 2009, ADAMS Accession No. ML093160816.

4) Letter from NRC to NSPM, "Summary of the December 15, 2009, Meeting to Discuss Technical Issues Related to a Future Application for Amendment Using MELLLA+ (TAC No. ME2449)," dated December 22, 2009, ADAMS Accession No. ML093451619.

Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM), hereby requests approval of an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) as described in Enclosure

1. The proposed change would allow Monticello Nuclear Generating Plant 2807 West County Road 75

Monticello MN 55362 S.00(

Document Control Desk Page 2 operation in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

domain. MNGP is currently licensed to operate in the Maximum Extended Load Line Limit Analysis (MELLLA) domain and with thermal-hydraulic stability Option I1. This proposed request for MELLLA+ expands the operating boundary without changing the maximum licensed core power, core flow, nor the current vessel dome pressure.

MELLLA+ requires changing the stability solution from Option III to the Detect and Suppress Solution -Confirmation Density (DSS-CD) solution and the use of the General Electric -Hitachi (GEH) analysis code TRACG02. This request includes the supporting TS changes necessary to implement the expanded operating domain, the change in stability solution, and the use of the TRACG02 analysis code. The terminology

."MELLLA+

operating domain" refers to all three of these items collectively.

The MELLLA+ operating domain expansion follows an EPU amendment request and the EPU conditions are then the bases conditions for the MELLLA+/- evaluations.

This bases assumption establishes the link and sequential nature of the EPU to MELLLA+relationship..

However, MNGP's EPU amendment is being delayed to allow the Nuclear Regulatory Commission (NRC) staff to develop additional review criteria for the Containment Accident Pressure (CAP) evaluation for the EPU application (Reference 1). A request for Concurrent review with the EPU amendment request was submitted in Reference 2 and was granted in Reference

3. The concurrent review facilitates future implementation plans and schedules without undue operational restrictions.

CAP is also evaluated for MELLLA+. The CAP evaluation confirms that the CAP issues in the MELLLA+ domain are within design limits and are bound by the EPU evaluation or are otherwise acceptable.

This is accomplished by determining the effects of MELLLA+ on the containment and suppression pool parameters.

Acceptable results of the CAP evaluation are shown in Enclosure 1, Attachment

4. There are no changes to the containment overpressure credit for MELLLA+. The NRC resolution of the CAP issue is expected in the spring of 2010. The effect on ECCS Net Positive Suction Head will be evaluated and submitted following receipt of NRC guidance.NSPM commits to resolve the CAP issue for MELLLA+ in the same manner as the issue is resolved for the delayed EPU amendment.

NRC approval of the requested operating domain expansionwould allow NSPM to implement operational changes to increase the operating flexibility of the plant due to the increased flow control range provided in the MELLLA+ operating domain. The MELLLA+ core operating domain does not require major plant hardware modifications.,The core operating domain expansion involves changes to the operating power/core flow map and changes to a small number of setpoints and alarms. Because there is no change in the operating pressure, power, steam flow rate, and feedwater flow rate, there is also no significant effect on the plant hardware outside of the Nuclear Steam Supply System (NSSS).

Document Control Desk Page 3 NSPM has evaluated the proposed changes in accordance with the requirements of 10 CFR 50.91 against the standards of 10 CFR 50.92 and has determined this request involves no significant hazard consideration.

The no significant hazard consideration can be considered complete although the CAP issue is unresolved.

This is because the evaluations demonstrate that the CAP issue is bound by the EPU peak loss-of-coolant-accident (LOCA) results (see Enclosure 1, Attachment 3, Sections 4.1 and 9.3.1).NSPM has committed to resolving CAP for MELLLA+ in the same manner as;EPU. If necessary, changes to the NSHC would be submitted at that time.The enclosure to this letter contains information supporting the proposed change. The enclosure and its attachments are described below.Enclosure 1 contains NSPM's evaluation of this proposed change. Included are a description of the proposed change, technical analysis of the change, regulatory safety analysis (No Significant Hazards Consideration and the applicable regulatory requirements/criteria), and an environmental consideration.

Attachment 1 provides a mark-up of the Technical Specifications indicating the proposed changes.Attachment 2 provides a copy of the associated draft mark-up TS Bases pages for information.

Attachment 3 contains the MELLLA+ safety analysis report (M+SAR). The M+SAR is an integrated summary of the results of the safety analysis and evaluations performed specifically for the MNGP MELLLA+ license amendment and follows the guidelines contained in General Electric (GE) Licensing Topical Report (LTR) NEDC-33006P-A; Revision 3, "Maximum Extended Load Line Limit Analysis Plus" (M+LTR). NRC has accepted use of this LTR for reference as a basis for a MELLLA+ license amendment request.Attachment 3 contains information which is proprietary to GEH. GEH requests that this proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390(a)4, as this is authorized by 10 CFR 9.17(a)4.

An affidavit supporting this request is provided in Attachment

4. A non-proprietary version of the M+SAR is provided as Attachment 5.Attachment 6 is a "Monticello MELLLA+ Risk Assessment," which provides an assessment of the MELLLA+ effects on risk relative to the current probabilistic risk assessment (PRA). This attachment supplements-M+SAR Section 10.5.NSPM plans to implement MELLLA+ following the 2011 refueling outage (RF025).Therefore, to support the NSPM schedule for the planned EPU power ascension which would occur following the completion of RF025, NSPM requests that the proposed Document Control Desk Page 4 amendment be approved by the end of the first quarter of 2011. Implementation of MELLLA+ is planned to be completed within .120 days from NRC approval of the.MELLLA+ LAR. In accordance with 10 CFR 50.91(b), a copy of this application, with non-proprietary attachments is being provided to the designated Minnesota Official.Summary of Commitments This letter makes one.new commitment and no revision to existing commitments:

NSPM will resolve the CAP issue for MELLLA+ in the same manner as the issue is resolved for the delayed EPU amendment.

I declare under penalty of perjury that the foregoing is true and correct.i Execu onJanu 21,2010.Ti .O'Connor Sit ice President Mo ticello Nuclear Generating Plant.Northern States Power Company -Minnesota cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce w/o proprietary attachments Enclosure with 6 attachments Enclosure 1 -to L-MT-10-003 NSPM Evaluation of Proposed Changes to Operating License and Technical Specifications for MELLLA Plus Enclosure I LICENSE AMENDMENT REQUEST MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (MELLLA+)EXPANDED OPERATING DOMAIN 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Operating License (OL) DPR-22 for Monticello Nuclear Generating Plant (MNGP). The proposed amendment includes supporting changes to the Operating License and Technical Specifications (TSs)necessary to implement the expanded operating domain and the change in stability solution required for MELLLA+.The proposed changes would change the TS to allow operation in the MELLLA+expanded domain, however, TS changes would be introduced that restrict operation in the MELLLA+ expanded operating domain if in single loop operation.

The changes will prohibit the use of the MELLLA+ expanded operating domain when in single loop operation.

Other changes include:*. Change to the allowable value (AV) for APRM-Simulated Thermal Power -High* Eliminating an unnecessary surveillance requirement

Changes in the Administrative TS to require certain content in the Core Operating Limits Report (COLR), and* Updating the applicable references in the administrative section.Nuclear Regulatory Commission (NRC) approval of the requested operating domain expansion will allow Northern States Power Company, a Minnesota corporation (NSPM)to implement operational changes to provide increased operational flexibility for power maneuvering, to compensate for fuel depletion,and to maintain efficient power distribution in the reactor core without the need for more frequent rod pattern changes.MELLLA+ increases the operating range to the Extended Power Uprate (EPU) Rated Thermal Power (RTP) at 80 percent flow; thus creating a 20 percent flow-control window. By operating in the MELLLA+ domain, a significantly lower number of control rod movements will be required than in the present operating domain. This represents a significant improvement in operating flexibility.

It also provides safer operation, because reducing the number of control rod manipulations: (a) minimizes the likelihood of fuel failures and (b) reduces the likelihood of accidents initiated by reactor maneuvers required to achieve an operating condition where control rods can be withdrawn.

Attachment 3 contains the MELLLA+ safety analysis report (M+SAR). The M+SAR follows the guidelines contained in General Electric (GE) Licensing Topical Report (LTR)NEDC-33006P-A, Revision 3, "Maximum Extended Load Line Limit Analysis Plus" (M+LTR) (Reference 1). The M+SAR provides the technical bases for this request and* contains an integrated summary of the results-of the underlying safety analyses and evaluations performed specifically for the MNGP expanded operating domain.The M+SAR also provides the analyses to change the MNGP stability solution from Option III to the Detect and Suppress Solution -Confirmation Density (DSS-CD) and the use of the General Electric -Hitachi (GEH) analysis code TRACG02. The DSS-CD stability solution is required by the M+ LTR Safety Evaluation Report (Reference

1) and is being implemented using the guidelines contained in General Electric (GE) Licensing Page 2 of 14 Enclosure 1 Topical Report (LTR) NEDC-33075P-A, Revision 6, "General Electric Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," (Reference 2). The use of TRACG02 is being implemented using the guidelines contained in GE LTR "DSS-CD TRACG Application," NEDE-33147P-A, Revision 2, November 2007 (Reference 3). The results of the DSS-CD evaluation and the use of TRACG02 are provided in Attachment 3, Section 2.4.Implementation of MELLLA+ is planned to be completed within 120 days from NRC approval of the MELLLA+ license amendment request.2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

The marked-up pages for the proposed changes to the Technical Specifications are included in Attachment 1.NSPM is requesting to implement the MELLLA+ expanded operating domain. This represents an operating domain of 2004 MWt (EPU RTP) with core flow as.low as 80%of rated core flow. The entire MELLLA+ domain is shown in Attachment 3, Figure 1-1.Proposed changes to the TSs are listed in Table 1 with a brief description of the basis for the change.NSPM proposes to make the supporting changes to the TS Bases in accordance with TS 5.5.9, "Technical Specifications (TS) Bases Control Program." Associated TS Bases changes are provided in Attachment 2 for information only. Attachment 3 provides retyped TS changes for convenience.

Page 3 of 14 Enclosure I Table 1 Monticello Proposed Operating License and Technical Specification Changes TS Section Description of Change Basis for Change TS 3.3.1.1 Reactor Protection System (RPS) Modify ACTIONs I, J, and K Attachment 3, Section 2.4.3 Instrumentation and to reflect new Backup and as described in-the DSSýStability Protection (BSP) CD LTR (Reference 2)ACTIONs I, J, and K requirements..

TS 3.3.1.1, '"Oscillation Power Range (Delete the Surveillance.)

Deleted to eliminate Monitor (OPRM) unnecessary actions..Instrumentation" Surveillance DSS-CD LTR (Reference 2).Requirements SR 3.1.1.1.16 Table 3.3.1.1-1, APRM Function 2b Revises the Allowable Value Attachment 3, Section 5.3.1 Simulated Thermal Power -High .to < 0.61W + 67.2% RTP.Add new note (h) to require Attachment 3, Section 2.4.3 ALLOWABLE VALUE. resetting the APRM-STP -High when the OPRM (Function 2f) is INOP.Table 3.3.1.1-1, APRM Function 2f Add note describing initial DSS-CD LTR (Reference 2), implementation arming Section 2.1 OPRM Upscale requirements.

TS 3.4.1, "Recirculation Loops Restrict MELLLA+ Operation MELLLA+ is not analyzed for in Single Recirculation Loop Single Loop Operation.

Operating" ' configuration. "Attachment 3, Section 3.6.3 LCO 3.4.1 TS 5.6.3 Core Operating Limits Report (COLR)a. Core operating limits shall be established Delete Item 6, the Adding DSS-CD LTR prior to each reload cycle, or prior to any requirement for the Period requirements (Reference 2)remaining portion of a reload cycle, and shall Based Detection Algorithm be documented in the COLR for the Setpoints and replace with: following:

6. The Manual Backup Stability Protection (BSP)Scram Region (Region I), The Manual BSP Controlled Entry Region (Region II), the modified APRM-STP setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1.b The analytical methods used to determine the core operating limits shall be those Replace Item 4, with: NEDO-Page 4 of 14 Enclosure 1 previously reviewed and approved by the NRC in the following documents:

33075-A, Revision 6,* "General Electric Boiling Water Reactor Detect And Suppress Solution-Confirmation Density Licensing Topical Report," January 2008 Delete Item(5) -NEDO-32465 is no longer applicable.

TS 5.6 Reporting Requirements Add reporting requirement DSS-CD LTR requirement 5.6.5, "OPRM Report" to (Reference 2)describe the reporting requirements required by TS 3.3.1.1, Condition I.Page 5 of 14 Enclosure 1

2.2 BACKGROUND

MNGP was originally licensed to operate at a maximum power level of 1,670 MWt.Northern States Power (NSP) has performed two previous power uprates. The first. uprate increased the licensed thermal power by approximately

6.3 percent

(References 4, 5, and 6). The second is an extended power uprate (EPU) currently under review (Reference 7)by the NRC, which increases the maximum power level to 2004 MWt.Operation of BWRs requires that reactivity balance be maintained to accommodate fuel burn-up. BWR operators have typically two options to maintain this reactivity balance: (a).control rod movements or (b) core flow adjustments.

Because of the strong void reactivity feedback and its distributed effect through the core, flow adjustments are the preferred reactivity control method. Operation at low-flow conditions at rated power level also increases the fuel capacity factor through spectral shift and the increased flow region compensates for reactivity reduction due to fuel depletion during the operating cycle.EPUs are implemented by extending the MELLLA operating, domain up to EPU power levels. The extension of MELLLA line to EPU power levels reduces the available core flow window. In addition, the increased core pressure drop with EPU limits the recirculation flow capability.

Consequently, EPU plants generally operate with a reduced core flow window and compensate for reactivity loss with control rod movement.

Operation at the MELLLA+ expanded operating domain will provide a larger core flow window for EPU plants..In June 2009, the NRC approved the use of the M+LTR (Reference

1) as a basis for MELLLA+ operating domain expansion license amendment requests, subject to limitations specified in the M+LTR and in the associated NRC safety evaluation.

The NSPM request complies with the specified limitations.

In January 2008, the NRC approved the use of the DSS-CD'LTR (Reference

2) as a basis for implementing DSS-CD as a stability solution to replace the Option Ill solution in license amendment requests, subject to limitations specified in the DSS-CD LTR and in the associated NRC safety evaluation.-

The NSPM request complies with the specified limitations.

The TRACG code for use in DSS-CD applications was approved by NRC in November 2007 (Reference 3). There is one limitation specified in the DSS-CD LTR SER and the NSPM request.comrplies with the specified limitation.

In addition, the NRC approved the Methods Licensing Topical Report (MLTR Reference 8)which imposes limitations and requirements for the use of GEH Methods in expanded operating domains including power uprates and MELLLA+ domains. The NSPM request complies with the specified limitations as discussed in Attachment 3.Detailed evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, and design basis accidents were performed and are provided in Attachment

3. These evaluations demonstrate that MNGP can safely operate in the MELLLA+ expanded operating domain with DSS-CD as the thermal hydraulic stability solution.Page 6 of 14 Enclosure I 3.0 TECHNICAL EVALUATION Attachment 3 summarizes the results of the significant safety evaluations performed that justify implementing the MELLLA+ expanded operating domain and the DSS-CD stability solution at MNGP. Attachment 3 is based on the M+LTR and the DSS-CD LTR. These evaluations demonstrate that MNGP can safely operate in the MELLLA+ expanded operating domain using the DSS-CD stability solution.Discussions of Issues Being Evaluated Plant performance and responses to hypothetical accidents and anticipated operational occurrences (transients) have been evaluated for the MELLLA+ license amendment.

The safety analysis (Attachment

3) summarizes the plant reactions to events evaluated for licensing and the potential effects on various margins of safety, and thereby concludes that no significant hazards consideration will be involved.MELLLA+ Analysis Basis The MELLLA+ safety analyses are mostly based on a Regulatory Guide (RG) 1.49 "Power Levels of Nuclear Power Plants," power factor times the rated power level. However, some analyses are performed at nominal rated power, either because the RG 1.49 power factor is already accounted for in the analysis methods or RG 1.49 does not apply.Fuel Thermal Limits No change is required in the mechanical fuel design to meet the plant licensing limits while operating in the MELLLA+ domain. No increase in allowable peak bundle power is needed and fuel thermal design limits will be met in the MELLLA+ domain. The analyses for each fuel reload are required to meet the criteria accepted by the NRC as specified in Reference 9 or otherwise approved by the NRC. In addition, future fuel designs will meet acceptance criteria approved by the NRC.Makeup Water Sources The BWR design concept includes a variety.of ways to pump water into the reactor vessel to deal with all types of events. There are numerous safety-related and non-safety related cooling water sources. The safety-related cooling water sources alone can maintain core integrity for.all postulated events by providing adequate cooling water. There are high and low pressure, high and low volume, safety and non-safety grade means of.delivering water to the vessel. These means include at least: " Feedwater (FW) and condensate system pumps* Low pressure core cooling system (Core Spray and Residual Heat Removal in Low Pressure Coolant Injection mode) pumps* High pressure coolant injection (HPCI) pump* Reactor core isolation cooling (RCIC) pump Standby liquid control (SLC) pumps* Control rod drive (CRD) pumps.Many of these diverse water supply means are redundant in both equipment and systems.The MELLLA+ operating domain does not result in an increase or decrease in the available water sources, nor does it change the selection of those assumed to function in the safety analyses.

NRC-approved methods were used to evaluate the performance of the Emergency Core Cooling Systems (ECCS) during postulated LOCAs.Page 7 of 14 Enclosure 1 Reactor Coolant Pressure Boundary The reactor pressure vessel (RPV) structure, RPV support components, and piping systems form the reactor coolant pressure boundary (RCPB) to contain reactor coolant and form a boundary against leakage of radioactive materials into the drywell. The RCPB was evaluated for theMELLLA+

operating domain conditions (pressure, temperature, flow, and radiation) and was found to meet their acceptance criteria for allowable stressesand overpressure margin.Design Basis Accidents (DBAs)DBAs are very low probability hypothetical events whose characteristics and consequences are used in the design of the plant, so that the plant can mitigate.their consequences to within acceptable regulatory limits. For BWR licensing evaluations, capability is demonstrated for coping with the range of hypothetical pipe break sizes in the largest recirculation, steam, and FW lines, a postulated break in one of the ECCS lines, and the most limiting small lines. This break range bounds the full spectrum of large and small, high and low energy line breaks; and demonstrates the ability of plant systems to mitigate the accidents while accommodating a single active equipment failure in addition to the postulated loss-of-coolant-accident (LOCA). Several of the significant licensing, assessments are based on the LOCA and include: Challenges to Fuel ECCS are described in Section 6.2 of the plant Updated Safety Analysis Report (USAR). The peak cladding temperature (PCT) calculated for a LOCA from the MELLLA+ domain is bounded by the license basis PCT that was calculated based on-rated flow. However, the ECCS performance evaluation (Attachment 3, Section 4.3) demonstrates significant margin to the criteria of 10 CFR' 50.46 and Appendix K at the reduced core flow of the MELLLA+ domain. Therefore, the ECCS safety margin is not affected by MELLLA+.* Challenges to the Containment

.-.Challenges to the Containment are described in USAR Section 5.2 wherein'the primary criteria of merit are the maximum containment pressure calculated during the course of the LOCA and maximum suppression pool temperaturefor long-term cooling.The peak values for containment pressure and temperature for events initiated in the MELLLA+ domain meet design requirements and confirm the suitability of the plant for operation in the MELLLA+ domain. The containment dynamic and structural loads for events initiated in the MELLLA+ domain continue to meet design requirements.

The change in short-term containment response is negligible" and, because there is no change in decay heat, there is no change in the long-term* response.

The containment pressure and temperature remains below the design* limits following any DBA. Therefore, the containment and its cooling systems are satisfactory for operation in the MELLLA+ domain.Design Basis Accident Radiological Consequences The magnitude of the potential radiological consequences depends on the quantity-of fission products released to the environment, the atmospheric dispersion factors, and the dose exposure pathways.

The atmospheric dispersion factors and the dose exposure pathways do not change. The quantity 6f activity released to the Page 8 of 14 Enclosure 1*environment is a product of the activity released from the core and the transport mechanisms between the core and the effluent release point. The radiological releases for events initiated in the MELLLA+ domain do not increase.The radiological consequences of LOCA inside containment, main steam line break, instrument line break, control rod drop and fuel handling accidents, are bounded by the evaluation at the current licensed operating domain and need not be reevaluated for the MELLLA+ domain. The radiological results for all accidents remain below the applicable regulatory limits for the plant.Anticipated Operational Occurrence (AOO) Analyses AOOs (i.e. transients) are evaluated to demonstrate consequences that meet the safety limit minimum critical power ratio (SLMCPR).

The SLMCPR is determined using NRC-approved methods. The limiting transients are core specific and are analyzed for each reload fuel cycle to meet the licensing acceptance criteria (Attachment 3, Section 2.2.1).Therefore, the margin of safety to the SLMCPR is not affected by operation in the MELLLA+ domain.Combined Effects DBAs are postulated using deterministic regulatory criteria to evaluate challenges to the fuel, containment, and off-site radiation dose limits. The off-site dose evaluation specified by RG 1.183 (Reference

10) and Standard Review Plan (SRP) 15.0.1 (Reference 11)provides more severe DBA radiological consequences scenario than the combined effects of the hypothetical LOCA. That is, the DBA, which produces the highest PCT and/or containment pressure, does not damage large amounts of fuel, and thus, the source terms and doses are much smaller.than those. postulated in RG 1.183 evaluations.

The conservatism of the combined effects is not reduced by operation in the .MELLLA+domain.Non-LOCA Radiological Release Accidents All of the limiting nonr-LOCA events discussed in USAR Chapter 14 were reviewed for the ,effect of MELLLA+. The dose consequences for all of the non-LOCA radiological release accident events are shown in Attachment 3, Section 9.0 to remain below regulatory limits.Equipment Qualification Plant equipment and instrumentation have been evaluated against the applicable criteria.The qualification envelope either does not change due to the MELLLA+ operating domain expansion or is bounded by the current licensed operating domain.Balance-of-Plant (BOP)The power, pressure, steam and FW flow rates, and FW temperature do not change for the MELLLA+ operating domain expansion; there are no changes to the BOP systems/equipment.

Probabilistic Risk Assessment Attachment 3, Section 10.5 and Attachment 6 describe the results of the Probabilistic Risk Assessments (PRAs) performed for MELLLA+. The best estimate of the Core Damage Frequency (CDF) risk increase for at-power internal events due to MELLLA+ is a delta CDF of 7.36E-8. The best estimate of the Large Early Release Frequency (LERF)increase for at-power internal events due to MELLLA+ is a delta LERF of 1.62E-8. Using the NRC guidelines established in Regulatory Guide 1.174 and the calculated results from Page 9 of 14 Enclosure 1 the Level 1 and 2 PRA, the best estimate.

for the CDF risk increase (7.36E-8/yr) and the best estimate for the LERF increase (1.62E-8/yr) are both within Region Ill (i.e. changes that represent very small risk changes).Environmental Consequences For operation in the MELLLA+ domain, the environmental effects will be controlled to the same limits as for the current operating power/flow map. None of the present environmental release limits are increased as a result of MELLLA+.There will be no change in the quantity of radioactivity released to the environment through liquid effluents-and no increase in airborne emissions of radioactivity as a result of MELLLA+. All off-site radiation doses will be small and within 10 CFR 20 and 10 CFR 50, Appendix I limits.As a result, it is concluded that the Monticello MELLLA+ operating domain expansion does.not constitute an unreviewed environmental question and is eligible for. categorical exclusion as provided by 10 CFR 51.22(c)(9).

Discussion of DSS-CD Stability Solution Issues For the Monticello MELLLA+, the long-term stability solution is being changed from the currently approved Option Ill solution to DSS-CD. The DSS-CD solution algorithm, licensing basis, and application procedures are generically described in NEDC-33075P-A (Reference

2) and NEDE-33147P-A (Reference 3), and are applicable to Monticello-including any limitations and conditions associated with their use and approval.The DSS-CD solution is designed to identify the power oscillation upon inception and..initiate control rodinsertion (scram) to terminate the oscillations prior to any significant-amplitude growth. DSS-CD is based on the same hardware design as Option Ill.However, it introduces an enhanced detection algorithm that detects the inception of power oscillations and generates an earlier power suppression trip signal exclusively based on.successive period confirmation recognition.

The existing Option III algorithms are retained (with generic setpoints) to provide defense-in-depth protection for ...unanticipated reactor instability events.Discussion of TSTF-493 Applicability There are no effects on the current TS or their licensing bases relative to TSTF-493.

Two TS Reactor Protection System (RPS) functions are changing in this amendment; the Oscillation Power Range Monitor (OPRM) and the Average Power Range Monitor (APRM)-Simulated Thermal Power (STP) High. The OPRM setpoints are unique to a particular core design for a particular fuel cycle. The OPRM function setpoints do not have specific TS allowable values. The APRM STP -High allowable values (AV) are specified in TS Table 3.3.1.1-1.

The power range neutron monitoring system amendment (TS Amendment 159) established that the allowable values and/or setpoints for these functions do not protect safety limits and are therefore, not subject to TSTF-493 requirements.

MELLLA+ changes the OPRM setpoints in that they are now from DSS-CD algorithms versus Option III algorithms, however, their, protective function remains the same and is not credited in the safety analysis.

Setpoint methodologies as described in TSTF-493 are not applied to these setpoints.

MELLLA+ also changes the APRM STP -High AV for two loop operations in the MELLLA+ domain and the APRM STP -High function is used for the Automated Backup Stability Protection (ABSP) if the OPRM becomes inoperable.

The APRM STP- High AV Page 10 of 14 Enclosure 1 and setpoint do have setpoint methodology applied as described in TSTF-493, however, the APRM STP -High function does not protect any safety limits whether in its normal anticipatory high power function or in the ABSP function.

The position that the ABSP does not protect a safety limit is described in the Technical Evaluation Report in the Safety Evaluation Report for the DSS-CD LTR (Reference 2). The APRM STP setpoints for the ABSP function are maintained in the COLR because they are unique to a particular core design for a particular fuel cycle as with the OPRM setpoints.

4.0 REGULATORY

SAFETY ANALYSIS 4.1 Applicable Regulatory Requirements 10 CFR 50.36 (c)(2)(ii)

Criterion 2,requires that TS Limiting Conditions for Operations (LCO) include process variables, design features, and operating restrictions that are initial conditions of design basis accident analysis.

Compliance with the TS ensures that the MNGP system. performance parameters are maintained within the values assumed in the safety analyses.

The TS changes are supported by the safety analyses and continue to provide a level of protection comparable to the current TS. Applicable regulatory requirements and significant safety evaluations performed in support of the proposed changes are described in Attachment 3.4.2, Precedent This application is submitted using the approved GEH license topical reports (References 1, 2, and 3) and their associated Safety Evaluation Reports. The MNGP application follows the methodologies and limitations of those LTRs and their respective SERs. This is the first application to request approval for the MELLLA+ operating domain expansion since:the referenced topical reports were NRC accepted.4.3 Significant Hazards Consideration

--Inaccordance with the requirements of 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM) requests an amendment for the use of the Maximum Extended Load Line Limit Analysis (MELLLA+)

operatingdomain for the Monticello Nuclear Generating Plant (MNGP). NSPM has evaluated the proposed amendment in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined that the operation of the facility in accordance with the proposed amendment presents no significant hazards. NSPM's evaluation against each of the criteria in 10 CFR 50.92 follows.1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?.

Response:

No.The probability (frequency of occurrence) of Design Basis Accidents occurring is not affected by the MELLLA+ operating domain, because MNGP continues to comply with the regulatory and design basis criteria established for plant equipment.

Further, a probabilistic risk assessment demonstrates that the calculated core damage frequencies do not significantly change due to the MELLLA+.Thereis no change in consequences of postulated accidents, when operating in the MELLLA+ operating domain compared to the operating domain previously evaluated.

The results of accident evaluations remain within the NRC approved acceptance limits.Page 11 of 14 Enclosure 1 The spectrum of postulated transients has been investigated and are shown to meet the plant's currently licensed regulatory criteria.

In the area of fuel and core design, for example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) is stillmet.Continued compliance with the SLMCPR will be confirmed on a cyclespecific basis consistent with the criteria accepted by the NRC.Challenges to the Reactor Coolant Pressure Boundary were evaluated for the MELLLA+ operating domain conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin.Challenges to the containment were evaluated and the containment and its associated cooling systems continue to meet the current licensing basis. The calculated post LOCA suppression pool temperature remains acceptable.

Therefore, the proposed changes do 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.Equipment that could be affected by the MELLLA+ operating domain has been evaluated.

No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified:

The full spectrum of accident considerations has been evaluated and no new or different kind of accident has been identified.

The MELLLA+ operating domain uses developed technology and applies it within the capabilities of existing plant safety-related equipment in accordance with the regulatory criteria (including NRC approved codes, standards and methods).

No new accident nor event precursor has been identified.

The-MNGP TS require revision to implement the MELLLA+ operating domain. The revisions have been assessed and it was determined that the proposed change will not introduce a different accident than that previously evaluated.

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

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

No.The MELLLA+ operating domain affects only design and operational margins.Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for the MELLLA+ operating domain conditions.

Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems and components, including the reactor coolant pressure boundary, will remain within their design allowables for design basis event categories.

No NRC acceptance criterion is exceeded.

Because the MNGP configuration and responses to transients and postulated accidents do not result in exceeding the presently approved NRC acceptance' limits, the proposed changes do not involve a significant reduction in a margin of safety.Page 12 of 14 Enclosure.

Based on .the considerations 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: ý(1) involve a significant

.!:increase in the probability or consequencesof an accident previously, evaluated; or (2)createthe 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.4.4 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would changean inspection or surveillance requirement.

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

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

Therefore,'

pursuant to.1 0 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. GE Nuclear Energy, "General Electric Boiling Water Reactor Maximum Extended Load. Line Limit Analysis Plus Licensing Topical Report," NEDC-33006P-A, Revision-3, Class III (Proprietary), June 2009 and NEDO-33006-A, Revision 3 Class I, (Non-proprietary), June 2009. -2..GE Nuclear Energy, "Detect And Suppress Solution-Confirmation Density Licensing Topical Report," NEDC-33075P-A, Revision 6, Class III, January 2008 and NEDO-33075-A, Revision 6, Class I (Non-proprietary)

January 2008.3. GE Nuclear Energy, "DSS-CD TRACG Application," NEDE-33147P-A, Revision 2, Class III (Proprietary), November 2007 and NEDO-33147-A, Revision 2, Class I (Non-proprietary)

November 2007.4. Northern States Power letter to NRC, July 26, 1996, License Amendment Request Supporting the Monticello Nuclear Generating Plant Power Rerate Program.5. Northern States Power letter to NRC, December 4, 1997, Revision 1 to License-Amendment Request Dated July 26 1996, Supporting the Monticello Nuclear Generating Plant Power Rerate Program.6. NRC Letter to Northern States Power, September 16, 1998, Monticello Nuclear Generating Plant -Issuance of Amendment Re: Power Uprate Program (TAC No.M96238).7. Xcel Energy Letter to NRC, November 5, 2008; License Amendment Request Extended.

Power Uprate (TAC MD9990).Page 13 of 14 Enclosure I 8.. GE Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains," Licensing Topical Report, NEDC-33173P, Class III, February 2006 and NEDO-33173, Class I, February 2006.9, GE Nuclear Energy, "General Electric Standard Application for Reactor Fuel," NEDE-24011 P-A and NEDE-2401 1 P-A-US, (latest approved revision).

10. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" July, 2000 ML003716792.

11. Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms,'" Revision 0, July, 2000 (ADAMS Accession No.MIL003721661).

12. Regulatory Guide 1.174, "An Approach for Using Probabilistic RiskAssessment In Risk-Informed Decisions On Plant-Specific Changes to The Licensing Basis." Revision1, April 2002.r Page 14 of 14 Attachment 1 of L-MT-1 0-003* Markups to the Technical Specifications for MELLLA Plus RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 4- AS requirod by Requir*ed TActie DP.1 and eAWL4 2e2.... .il l lLnt.ato ILternate mot.od to detect and eupperoe!SGBF~Rqa Y6A--19 96601 NOTEG 4 2 I;-, , s See Insert I 120 da'ys for Actions I, J and K.RAstAm rý "*r^d GhARRAIS to OPERAB LE=J. Required Action and 44 Reduce THERMAL 4-heer-asseciatcd Completion POWER to =04 RTP.Tfime oef Con~dition SURVEILLANCE REQUIREMENTS

NOTES --------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.-2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2

NOTE ----------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER ! 25% RTP.Verify the absolute difference between the average 7 days power range monitor (APRM) channels and the calculated power is < 2% RTP while operating at25% RTP.Monticello 3.3.1.1-3 Amendment No. 4-49,4-W I Insert 1 I. As required by Required Action D.1 and referenced in Table 3.3.1.1-1.

I.1 AND T931 Initiate action to implement the Manual BSP Regions defined in the COLea..e.a.e-ffithed te deteet and Tmuppress thermal hydratilic instability Tmnl smmsnt- thim Automated BSP Scram Region using the modified APRM Simulated Thermal Power -High scram setpoints defined in the COLR.AND 1.2.2 Initiate action in accordance with Specification 5.6.5.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 90 days 4 4 J. Required Action and associated Completion Time of Condition I not met.J.1 Initiate action to implement the Manual BSP Regions defined in the COLR.AND J.2 Reduce oDeration to below Immediately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 120 days the BSP Boundary defined in the COLR.AND J.3 -NOTE LCO 3.0.4 is not applicable Restore required channel to OPERABLE 4K. Required Action and 4K.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to <20% RTP.Time of Condition not met.

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.12 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 Verify Turbine Stop Valve -Closure and Turbine 24 months Control Valve Fast Closure, Acceleration Relay Oil Pressure -Low Functions are not bypassed when THERMAL POWER is > 45% RTP.SR 3.3.1.1.14

NOTES ----------------

1. For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Testing of APRM and OPRM outputs shall alternate.

2. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS SR 3.3.1.1.15

NOTES ----------------

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1 Vo14 re.ify the ....cilltion powor ..ng. oitr (OPRM), 24- meh function is not bypPssedswhen -A- 259 PRR r 'JAtAnm '--- latoe d IDELETE= Thrý oo t~25 PPdroircuAtion drive flow et 60% o-f rate roircul1AtionA drFiVe flo-w.Monticello 3.3.1.1 -5 Amendment No. -46, 159 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 SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 .REQUIREMENTS VALUE 1. Intermediate Range Monitors a. Neutron Flux -High High 2 3 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 3.3.1.1.14 122/125 divisions of full scale< 122/125 divisions of full scale 3 b. Inop 2 3 5 (uI 3 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 3.3.1.1.4 SR 3.3.1.1.12 0 SE Z31.1.1 2. Average Power Range Monitors a. Neutron Flux- High, (Setdown)2 3(c)b. Simulated Thermal 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 from a core cell containing one or more fuel assemblies.(b)RTP when reset for single loop operation per LCO 3.4.1, "Reclrculatlon Loops c 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 Retyped to Reflect PRNMS I (h) With OPRM Upscale (function 2.f) inoperable, reset the APRM-STP High scram setpoint to the values defined by the COLR to implement the Automated BSP Scram Region in accordance with Action I of this Specification.

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.1 REQUIREMENTS VALUE c. Neutron Flux -High d. Inop.e. 2-Out-Of-4 Voter 1 1,2 1,2 3 (c)3 (c)2 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(f(g)SR 3.3.1.1.15 G SR 3.3.1.1.4 SR 3.3.1.1.15 G SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

< 122% RTP NA NA f. OPRM Upscale(e)

3. Reactor Vessel Steam Dome Pressure -High> 20% RTP 3(c)1,2 2 I 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 3.3.1.1.15R331 4 G SR 3.3.1.1.4 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.14 G SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 S R 3.3.1.1.12 S R 3.3.1.1.14 As 6pecified

[NA! Deletel< 1075 psig 4. Reactor Vessel Water Level -Low 1,2 2> 7 inches (c) Each APRM / OPRM channel provides inputs to both trip systems. IAdd new note (e) from Insert 2 here (e)DUuron wne ui-'-1' monieorin I'Reioa the GIiRRM -pcI ---- io I opork'=a (f) If the as-found channel setpoint is not the Nominal Trip Setpoint 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, the 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. 44&, 159 Insert 2 (e) Following DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region.However, DSS-CD is considered OPERABLE and shall be maintained OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region.

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 LCO 3.4.1 Recirculation Loops Operating Two recirculation loops with matched flows shall be in operation.

OR II f nn i md..o r....c.latin.

loop sall. bo i operatlon prGoVie, te n fg m.t.One recirculation loop may be in operation provided the plant is not operating in the MELLLA+domain defined in the COLR and provided the following limits are applied when the associated LCO is applicable

/1 ar; ap,..lied

,,henm the a.....atod L-CO is applicablo:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitor Simulated Thermal Power -High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY:

MODES 1 and 2.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the requirements of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO not met. the LCO.B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.OR No recirculation loops in operation.

Monticello 3.4.1-1 Amendment No. 41-46, 159 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 3.3.2.1-1 Rod Block Monitor Functions 1 .a, 1 .b, and 1 .c and associated Applicability RTP levels;5. 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 6. Ro...t.r Protoction Systom lrctr'ment1tion P riod B.-cod Dotcction See Insert 3 for FAlgorithm -trp c ---t a....oc.iatod Yith Tabo 3.1.1 1, Function 24., inew Item 6. 0-6RM-U.peeaJe

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 1-P-A, "General Electric Standard Application for Reactor Fuel";2. NSPNAD-8608-A, "Reload Safety Evaluation Methods for Application to the Monticello Nuclear Generating Plant";3. NSPNAD-8609-A, "Qualification of Reactor Physics Methods for Application to Monticello";

4- NEDQ 31690, ",,VVR OWn.r.' Group Long Term Stability 6c'uti8n See Insert 3 for Licen.ing Methodology";

..d new Item 4.n t N EDO 3212695 A, "R,,o.,t Stability Detect and Suppe.. SE..Olut LcnigBasis Mothodology and Rolead Applicationc,2 Auguet 1006.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, 159 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Post Accident Monitorinq Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.5.6.5 OPRM Report When a report is required by CONDITION I of LCO 3.3.1.1, "Reactor Protective System (RPS) Instrumentation," a report shall be submitted within 90 days of entering CONDITION I. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.Monticello 5.6-3 Amendment No. 4-46, 159 Insert 3 5.6.3.a 6. The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power -High setpoints used in the OPRM (Function 2f), Automated BSP Scram Region, and the BSP*Boundary for Specification 3.3.1.1.5.6.3.b 4. NEDO-33075-A, Revision 6, "General Electric Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," January 2008 Attachment 2 of L-MT-1 0-003 Markups to the Technical Specifications Bases for MELLLA Plus Information Only RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) more of the other APRM Functions through that voter is still maintained.

This may be considered when determining the condition of other APRM Functions resulting from partial inoperability of the 2-Out-Of-4 Voter Function 2.e.There is no Allowable Value for this Function.2.f. Oscillation Power Range Monitor (OPRM Upscale)lReplace with Insert 4 The OPIRM Upscale Function provides comnpliance with QDrC 10 and GDC 42, t!ho-oby proViding prt."ti .... ,-n from ecding....,;.

the fuel MPR safcty limit (SL) duo to anticipatod thorm.al hydraulic pwer Reforoncos 8, 4 9n,and 20 describo.

three algorithms; detocting thermal hydraulic0G intbltFelatod neutron flu oscillations:

the porFiod bAccd detoct"*iEn

.the amplitude bacod algorithm. , and the frcwth Fate algorithm.

All three a^. implemented in the QPRM Upscalc Function, but thc saf...' analysis takes credit on.ly for the p .od based dctctin algorithm.

The remaining algorithmRS provigde defence in depth andl additieRal protection against unanticipated PRM UpscalI Function OPERA.I LITY fr Technical Specification is ba;ed only on the period based detection1 algoritm.The OPRM Upscale Functfion recneives in~put signals from the local poWer ran~ge monRitors (L=PRMS) Within the reactor core, which are combined into"ecols" for evaluation by the OPRM algorithmc.

T-he OPRM Upscale F=unctio is reuie to beO RABL F when tRh plant is at ýt 20%0 RTP, the reio of power flew operation where anticipated 8eventS could0 le-ade to hlermýAal hydraulic instability and related neutro-n flx O.cillations.

^;':th* .thi region, the in..M.Ups.al

.Functien is automatically ea.bled (bypass removed) wheni THERMAL POWER, as indicated by the; APRA Simulated Th ermal Power, ic 4 RT-P RRd reactor core flow, as inice by reGic.ati.

n drive flow, ..60% o0 rated flow, the operating regioni where actual therm~al hydraulic instability and related neutFro flux oscillations Ma occ)Gur. The loere bound, 20%RTP, ir choen to proide margin in the unlikely event of a reacGte power inraetransient occu-rring while the plant is 9opeating below the 25%4 automa-tc OPRM Upscale rip "e"able point.An UP pscale tiP sued fRo an RA channelwAhen the p~e An O£R.M i cai trip !s;...' .. ADRM periodl .k^ k .. ^a ed detectiAon aI9goithmin that Rhannel detect; osGc1iator G;, hangesin the neutron flux(, indicated by the combined signals of the ILPRMA detectarsi in a cell, With period confirMmQatin and relativeA cell amplitude exceeding, the neuAtqro flux, indicated by the combinedl 6ignale of the L-PRM. d-etectrsM-inR a cell, with period con..AfiRrmatin anRd re-lativet cell1 amplitude exceedinig Monticello B 3.3.1.1-13 Revision No. 12 Monticello B 3.3.1.1-13 Revision No. 12 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Cpecified cetpeinte.

One or more coIc in A chAnnel eXceoding tho trip condtf.i..

will ;roult in -A c han cri AR pcctalo trip ic ....icudfro9m the- c-h;annel-if oither: the growA.th rate or -amplitude base al, ithms dotect gri owiniocillator, changoc in the n, nunn fl uxI for one or more col1c i that channel Threq tho four chRn;e,4 e rie be &aRnei capable of detecting therlF.

i"tabilities, by detetiRng the FolAtod neutAro fluX ocltmens, and iccuing a trip cignal before tc MCrPR SL1i excoodod.no-minal;;

seftings determfined applying the ctability analycic, licnscin meth"odolgy (Rfc.m 18, 159 and 20) developed by the UpWVR Ownctc Group and- CGorVessal Eletric. Theresn l e Value for thHFuto Ti1h Anttincr are not tradeitoa piresumetatie n co mptpit detrmned under An IthMens t cetpoisnt mnethdeology.

Since the settinge may vaTi cayel to cycle, the allowable value oElumn in Table 3.3.1.1 1 indicatcz the OrRM Uprceale Func~ition trip settinga, i.e., the Pcriod based dctcction a@lgorithmf, are "Ant tpoifnead whin R" challenge the C R.I crdance with the N fRC R cluating frA tAendeRnPt 15N (Ref. 21), the OPRM Upkale Funirctitfon net LSS SL2 related.3. Reactor Vessel Steam Dome Pressure -High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion.

This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function.

However, the Reactor Vessel Steam Dome Pressure -High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 9, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Neutron Flux 7 High signal, not the Reactor Vessel Steam Dome Pressure -High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.High reactor pressure signals are initiated from four pressure switches that sense reactor pressure.

The Reactor Vessel Steam Dome Pressure-High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.Monticello B 3.3.1.1-14 Revision No. 12 L Inser 41 2.f. Oscillation Power Range Monitor (OPRM) Upscale The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.

Reference 18 describes the Detect and Suppress -Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA). aný-Reference 18 also describes the DSS-CD Armed Region and thef....r.ne..

18, 19 and 20 describe three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period based detection algorithm (PBDA), the amplitude based algorithm__.ApA, and the growth rate algorithm_(GRA). All th-eefour algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period bascd detection algorithmCDA.

The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY f.r T..hni.al pe..ifi.atie.

prpooses is based only on the period based detectien algorithmCDA.

The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.

DSS-CD operability requires at least 8 responsive OPRM cells per channel.The OPRM Upscale Function is required to be OPERABLE when the plant is eat--- 20% RTP, which is established as a power level that is greater than or equal to 5% below the lower boundary of the Armed Region. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability d related neutron flux oscillations.

Within this rogion, ale Function is automatically trip-enabled THE WER, as indicated by the APRM Simulated Thermal Power, is -RTP corresponding to the plant-specific MCPR monitoring threshold and reactor core flow, as by recirculation drive flow, is < less than 75%- of rated flow., the oporating region where actual thermal hydraulic eseillatiens and related n.utr.n fiux acs.illatien ffay-eeeid.

The lewor- bouand, 290e RTP, is chesen to provido margin iR the unlikely event of a reacteir po;;e .+/-nras pewer 47nerease trans-tent eeeurrmng w+/--he ttne jq~ase :t eperat-ng tietw the 2S% autom.ati.

PE Upocale trip. onale point. This region is the OPRM Armed Region. Note e allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing though the DSS-CD Armed Region during both a single startup and a single shutdown following DSS-CD implementation.

Note e reflects the need for plant data collection in order to test the DSS-CD equipment.

Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.

I insert 4 continued I An OPRM Upscale trip is issued from an A4-OPRM channel when the period based d.t..tieonconfirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by period confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. the .ombined signals of the EP4l d.t..t.......

a eel!, with peried eenfirfmotiens and rolotivo1 ad mplitude emcooding specified setpeints.

Ono er rre:ro cobl in a channe! e3Ecoodingf the trip conditiens will rozuilt in a channol trip. -An OPRI'Upscale trip is also issued from the channel if either any of the defense-in-depth algorithms (PBDA, ABA, GRA) heh rew-th rate eir amplitupo based algerctn sc doteet growing z1 changcs in the noutron fluxexceed t-heits trip condition for one or more cells in that channel.Three of the four channels are required to be operable.

Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded.

There is no allowable value for this function.The OPRM Upscale function settings are not traditional instrumentation setpoints determined under an instrument setpoint methodology.

In accordance with the NRC Safety Evaluation for Amendment 159 (Ref. 24), fh= )PPRA I Iorel Iiin Pr inn ic nnf I £Q Z 41 -rmilfa rI 4 nr oi nron 9A confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology.

RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

D. 1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1.

The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed.

Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.

E.1, F.1, G.1 and J.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Actions E.1 and J.1 are consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)." H.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted.

Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.1.1 if ,PRMI VU p al trip bi.it; -' no9t mainta*R9nd, Con.ditfion 1I extc.Reference 17 justified use of alternate metheds te detect and suppress Replace with Insert oscillAtoncs-for a limited p-iod of tim". The alternate m1thod6 a.o 5 for Bases of procedurally ectabliched ten. t with the guid"linoc

dontifi1 in Action 1.1 Rofero.co 22 requr,.ing .manual operator ac.tion to9 ra. the plant i Mctontiel B. 31..-4RviinN.1 Monticello B 3.3.1.1-24 Revision No. 12 RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)

See Insert 5 for 1.2 INote moved to J.3 I ertain p.Vdefinedl eventg. occur. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ComRpl.tion Timf.e i d on ennoin judgment to Rk4allo orderi' tr#aneitfion to the altornlatc mothod whilo,, ,,lim"iting the piod of time during Which no Automa*tic altoAFRnaRtoR dotect and suppress trip capability is foFrmally inplaco. Based on the small prob9ability of an instability evenAt ocrigat all, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time ic. judged to be roaso able 44 The altcrnato method to dotect and su1ppr86ess sillations implcmcntod in accort;dance, with Required ctionR 1.1 was evaluated (RofcFRonc 1:7) based on use for Up to 120 days only. The evaluation, based on enginccrin judgm~ent, conclu-deRd-that the liktlihooAd-of an intbltevent that could-not be adequately handled by the alternate-mfethod-s during this, 120 day period was negligibly small. The 120 day period ic intcnded to be an outside limit to alloW for the c-ase vohere design changcs or x~ cniv analysis might be roqUircd to understand-or coArrecrt come unaniticipatcd chaactrisic f the italtydctoction algorithmsG or equipmgent This actionA ifs not intended and was niet evaluated acs a rouiA-w_.

altcPrnat;ýivc to returning failed or inoporablc equipment to OPERABLEstus Corr~ection of ro)utine equipm~ent failure Or inprbliyi xpcctc t noFrmally be accomplishedd wthnthe com~pletion; timer, allowcd for the cstiens for Conditions A and B3.A Note is prov.idcPed to indicate;RR tha~t CO3.0.4 Is noet applicable.

The intcnt o-f thaRt note is to allow plant startup while operating wihnthe 1:20a Completion Time for Required Action .2 Th piar'purpso ofti*63uio is t allow-I An orderly comApletion of design an~d Verification activities, in the event of a required design cshange, without un~duc impacst on plant operation.

See insert 5 for Bases of Actions J and K.Monticello B 3.3.1.1-25 Revision No. 12 Monticello B 3.3.1.1-25' Revision No. 12 Insert 5 ACTIONS I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required.Reference 17 jusctifid use af

... t- detat- and suppress oscillations fer a li-ited peried -tifn. -The ifiManual BSP Regions are described in Reference

18. The alternatemManual BSP Regions methads arc are procedurally established consistent with the guidelines identified in Reference a-L-18 and requir4ei specified manual operator actionsactian ta seram the plant if certain predefined events operational conditions occur.The Completion Time of immediate is based on the importance of The 12 heur Campla ti Tý6ffie. is base d en engineevtiag jtdgfaent te allow arderly transitian ta the alternate maetheds while limiting the period of time during which no automatic or alternate detect and suppress trip capability is feorally in place. Basad an the small prebability af an instability event eeeaurriing at all, the 12 haur Campletian Time is judged ta be reasenable.

1.2.1 and 1.2.2 Actions 1.2.1 and 1.2.2 are both required to be taken in conjunction with Action I.1 if OPRM Upscale trip capability is not maintained.

As described in Section 7.4 of Reference 18, the Automated BSP Scram Region is designed to avoid reactor F instability by automatically preventing entry into the region S 0 ower and flow-operating map that is susceptible to reactor ins a The reactor tri would be initiated b the modified APRM f1--bia d scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection.

The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal hydraulic instability events.-Backup Stability Protection is intended as a temporary means to protecti.Qa against thermal-hydraulic instability events.The reporting requirements of Specification

5.6.5 document

the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.J.1 If the Required Actions I are not completed within the associated Completion Times, then Action J is required.

The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.

J.2 The BSP Boundary, as described in Section 7.3 of Reference 18, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that the immediate final statepoint for a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion.

Potential instabilities would develop slowly as a result of the feedwater temperature transient (Reference 18).The Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to complete the specified.-t- 4 nrc .r cn hi c -. -., based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant system.J.3 Backup Stability Protection (BSP) is a temporary means for protection asainst thermal-hydraulic instability events.An extended period of inoperability without automatic trip capability is not lustified.

Consequently, the required channels are required to be restored to OPERABLE status within 120 days.Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Action J.l) and the BSP Boundary (See J.2) during a 120-day period is negligibly small. The 120-day ner-jod is intended to allow for the case where limited.....a n e o i s .......... ... ........... ..... ..... ... .......... ..A-cs -n nhn1a, -c-r- -- c, 4 -rI ¶YC4 --j' Iu-, h I r-r, 4 -A i-rn understand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

This action is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status.Correction of routine equipment failure or inoperability is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and B K. 1 If the required channels are not restored to OPERABLE status and the Required Actions of J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least 20%RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is Irea.sonahi hased on onerat~na exoeriencee to reach the Isnecified oneratina ower level from full nower conditions in an orderly manner and without challenging plant systems.K Moved from 1.2 A Note is provided to indicate that LCO 3.0.4 is not applicable.

The intent of that note is to allow plant startup while operating within the 120-day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. The 24 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.1.1.15 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels.

The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing

-- applicable to Function 2.b and 2.f only), the 2-out-of-4 voter channels, and the interface connections into the RPS trip systems from the voter channels.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The 184 day Frequency of SR 3.3.1.1.15 is based on the reliability analysis of References 17 and 21. (NOTE: The actual voting logic of the 2-Out-Of-4 Voter Function is tested as part of SR 3.3.1.1.12.)

Note 1 is provided for Function 2.a to clarify that this SR is required to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. Note 1 allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.Note 2 is added to clarify that the CHANNEL FUNCTIONAL TEST is limited to the recirculation flow input processing and does not include the flow transmitters.

SR 3.3.1.1.146 Thic SR nc.ureo that scra.,mc .initi=ted from OPRM Upccalo Function (Function 2., will inot be inadvotently bypaccod when the THERMAL POWER, as i 'dicated by thoeAPRM Simulated Thrm-al Power, -c tca 26%RT-P and coro flow, ac indicated by rocirculation drive flow, ic 19 60% Fated corP8 flow.A Thic nrmF~ally involvoc98-cofrm,-RFing th9 bypace coetpeints.,Adequate marginc for the intumn etpoint mnethodologiec aro incorporated into the actual retpowint.

The acualwueillanco oncuroc that the G PR A. U-psrcalo Func~qtion.

ic enab-le-d (not bypaccod) for the corrcct.alu-eqs of A.PRMN. Simulat-d-Ther9mKal PoWer- -and- Arecircul11ation drive flew.Monticello B 3.3.1.1-34 Revision No. 12 Monticello B 3.3.1.1-34 Revision No. 12 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 331111and the Monticello core flow mneasuom~ent 6yctomf calibration procod-ur ncro-W that the APRM Simulatod Thermal PW'cr ,nd reci.cuat.on floW prop.rlY -orrelatA with THERMA.L POWER and cr.9e floW, rocpectively.

if any bypass cetpoint non c...R.ativ. (i.e., the OPRM'.p..a.

Function io bypasted when ADPR Simulated Thermal Powcr .. 25 and reiclaind-rive fleow !- 690% rated), thonR the_ affecrte~d channel i considcrcd inoperable for the OPRM Uptc9ale Function.

Altcrnlativcly, the byparr, retpoint may be adjusted to place; the channel1 in a conser~atiVe condition (non bypass). if placed in the non bypass condition, this SR is maet aRd the chanRnR ic ccRRcldcd OPERABLE.The Frcequency of 24 moneths it based oneqnern judgmcnit an reliability of the compenents.

The Frequency of SR 3.3. 1. 1.16- for the APRAM OP2RM Upealo FuncRtion ithated pna2 mneth calibraio Fintcal (Refs. 17 and 21).REFERENCES

1. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation." 2. USAR, Section 7.6.1.2.1.

3. USAR, Section 7.6.1.2.5.

4. USAR, Chapter 14.5. USAR, Chapter 14A.6. USAR, Section 7.8.2.1.7. USAR, Section 7.3.4.3.8. Not Used.9. USAR, Section 14.5.1.10. USAR, Section 14.7.1.11. USAR, Section 14.7.2.12. USAR, Section 14.7.3.Monticello B 3.3.1.1-35 Revision No. 12 Monticello B 3.3.1.1-35 Revision No. 12 RPS Instrumentation B 3.3.1.1 BASES REFERENCES (continued)

13. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.14. USAR, Section 14.4.5.15. USAR, Section 14.4.1.16. NEDC-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.17. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function", October 1995.See Insert 6 for updated References.

NEDO-PQ 319-60 A, "BVVR Ownorc' Group Long TcrmF Stability Solutin Liccning Mthodlo...

," .....bo 100 95.NEDO 31060 A, Supplomont 1, "BWVR Qwnorc' Grcup Long TcFRm Stability So.lutfios

_iccnsing Mothodolegy," ..v.mbc. 1005.NEDO A, "B,^R Q,' .....R..' Group Long T"rm Stability A __:-:- --....1 A ._.-A 4 AAti-ensine aaRd I le.Aead Fd b , f %W q U bt 21. NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip. Function", November 1997.22- 6.-A.England .WWUGU) letter to M. .VIrgillo (WRn), "BWV WnoR..v * ,-Group GuJdolinos for Stability Interim Corroctivo A.cton," 23. U.S. NRC Regulatory Issue Summary 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006.24. Amendment No. 159, "Issuance of Amendment Re: Request to Install Power Range Neutron Monitoring System," dated February 3, 2009. (ADAMS Accession No. ML083440681)

25. GHNE-0000-0073-4167-R2, "Reactor Long-Term Stability Solution Option IIl: Licensing Basis Hot Channel Oscillation Magnitude for Monticello Nuclear Generating Plant," December 2007.26. Letter from GEH to NRC, "NEDC-33075P-A, Detect and Suppress Solution -Confirmation Density (DSS-CD) Analytical Limit (TAC No. MD0277)," dated October 29, 2008.Monticello B 3.3.1.1-36

-Last Revision No. 12 Monticello B 3.3.1.1-36

-Last Revision No. 12 FInsert 6fl 18. NEDC-33075P-A, Revision 6, "General Electric Boiling Water Reactor Detect and Suppress Solution -Confirmation Density," January 2008 19 Not used 20. Not used 21. NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function", November 1997.22. Not used N The Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

operating domain is not analyzed for single Recirculation Loops Operating recirculation loop operation, and therefore can not be B 3.4.1 utilized in single recirculation loop operation. (Ref. 8)BASES APPLICABLE SAFETY ANALYSES (continued)

System (RPS) average power range monitor (APR ) Allowable Values is also required to account for the different relationshi s between recirculation drive flow and reactor core flow. The A LHGR and MCPR limits for single loop operation are specified in the C LR. The APRM Simulated Thermal Power -High Allowable Value is t LCO 3.3.1.1,"Reactor Protection System (RPS) Instrumentation." Recirculation Loops Operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO Two recirculation loops are normally in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation.

With only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2,"MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Simulated Thermal Power- High Allowable Value (LCO 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of Reference 3.APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 With the requirements of the LCO not met the recirculation loops must be restored to operation with matched flows within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation.

Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses.

Therefore, only a limited time is allowed to restore the loop to operating status.Monticello B 3.4.1-3 Revision No. 12 Recirculation Loops Operating B 3.4.1 BASES ACTIONS (continued)

Alternatively, if the single loop requirements of the LCO are applied to the APLHGR and MCPR operating limits and RPS Allowable Value, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump.BI With no recirculation loops in operation or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.

The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch.

At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit and the APLHGR requirements reduce the average planar bundle power such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The jet pump loop flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. The mismatch is measured in terms of percent of rated core flow. If the flow mismatch Monticello B 3.4.1-4 Revision No. 12 Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE SR 3.4.1.1 (continued)

REQUIREMENTS exceeds the specified limits, the loop with the lower flow is considered not in operation.

This SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation.

The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loops are in operation.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.REFERENCES

1. USAR, Section 14.7.2.2. USAR, Chapter 14.3. NEDC-32514P, "Monticello SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," October 1997.4. NEDO-24271, "Monticello Nuclear Generating Plant Single-Loop Operation," June 1980.5. NEDC-30492, "Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Power Generating Plant," April 1984.6. NEDC-32546P, "Power Rerate Safety Analysis Report for Monticello," Revision 1, July 1996.7. USAR, Section 14.6.\ USARyssPlLn "Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 3, June 2009.Monticello B 3.4.1-5 -Last Revision No.12 Attachment 4 of L-MT-1 0-003 MELLLA Plus Safety Analysis Report Affidavit GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Edward D. Schrull, state as follows: (1) I am Vice President, Regulatory Affairs, Services Licensing, GE-Hitachi Nuclear Energy Americas LLC ("GEH"). I have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in NEDC-33435P, Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus, Revision 1, dated December 2009. The proprietary information in NEDC-33435P is identified by a dark red font and dotted underline placed within double square brackets, This sentence is an ex.ample..

In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).(4) Some examples of categories of information which fit into the definition of proprietary information are: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Affidavit for NEDC-33435P Revision 1 Affidavit Page I of 3 GE-Hitachi Nuclear Energy Americas LLC The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a"need to know" basis.(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results and conclusions regarding supporting evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of the Maximum Extended Load Line Limit Analysis Plus analysis for a GEH Boiling Water Reactor ("BWR"). The analysis utilized analytical models and methods, including computer codes, which GEH has developed, obtained NRC approval of, and applied to perform evaluations of Maximum Extended Load Line Limit Analysis Plus for a GEH BWR.The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities.

The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.The value of the technology base goes beyond the extensive physical database and Affidavit for NEDC-33435P Revision 1 Affidavit Page 2 of 3 GE-Hitachi Nuclear Energy Americas LLC analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very.valuable analytical tools.I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.Executed on this 16th day of December 2009.Edward D. Schrull Vice President, Regulatory Affairs Services Licensing GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd.Wilmington, NC 28401 edward.schrull@ge.com Affidavit for NEDC-33435P Revision 1 Affidavit Page 3 of 3

v t e Letter Sequence Other
TAC:MD9990, Clarify Application of Setpoint Methodology for LSSS Functions (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 16 December 2008 11 February 2009 11 March 2009 12 March 2009 18 March 2009 18 March 2009 19 March 2009 20 March 2009 23 March 2009 29 March 2009 28 March 2009 28 March 2009 6 April 2009 6 April 2009 29 April 2009 26 June 2009 2 July 2009 2 July 2009 14 July 2009 14 July 2009 5 November 2009 16 March 2011 12 September 2011 4 October 2012 8 November 2012 28 March 2013 17 May 2013 9 July 2013 26 November 2013 18 December 2008 18 December 2008 19 March 2009 25 October 2009 15 February 2013 12 March 2013 10 May 2013 24 April 2013 Answers 25 January 2010 28 September 2010 13 January 2012 21 December 2012 7 March 2013 18 March 2013 21 March 2013 10 April 2013 18 July 2013 18 July 2013 2 August 2013 18 May 2016 19 July 2012 31 January 2013 29 March 2013 10 April 2013 Results Approval, Approval, Approval, Approval, Approval, Approval Other: L-MT-08-091, Calculation 0801040.301, Steam Dryer Outer Hood Submodel Analysis., L-MT-09-002, Response to NRC Probabilistic Risk Assessment (PRA) Branch Requests for Additional Information (Rais) Dated December 5, 2008, L-MT-09-003, Response to NRC Environmental Branch Requests for Additional Information (Rais) Dated December 18, 2008, L-MT-09-004, Enclosure 1 (Continued) to L-MT-09-004, NSPM Response to Containment & Ventilation Branch RAI Number 2, Copy of MNGP Appr T0406 Data Transmittal Pages 1 of 538 Through Pages 538 of 538, L-MT-09-026, Calculation 0000-0081-6958 MNGP-PRNMS-APRM Calc-2008-NP, Rev. 1, Average Power Range Monitor Selected Prnm Licensing Setpoints - EPU Operation (Numac), L-MT-09-027, Extended Power Uprate: Response to Instrumentation and Controls Branch RAI No. 3 Dated April 6, 2009, L-MT-09-044, Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (Emcb) Requests for Additional Information (Rais) Dated March 28, 2009, L-MT-09-049, Drawing C.5-2007, Revision 15, Failure to Scram, L-MT-09-073, Extended Power Uprate: Response to NRC Containment and Ventilation Review Branch (Scvb) Requests for Additional Information (Rais) Dated July 2, 2009 and July 14, 2009, L-MT-09-083, Extended Power Uprate: Limit Curves Requested by the Mechanical and Civil Review Branch (Emcb) Associated with Requests for Additional Information (Rais) Dated March 20, 2009, L-MT-09-088, Extended Power Uprate: Revision to Clarify Text in Enclosures 5 and 7 of L-MT-08-052, L-MT-09-097, Extended Power Uprate: Acknowledgement of NRC Review Delay, L-MT-10-025, Extended Power Uprate (Epu): Response to NRC-NSPM February 25, 2010 Conference Call, L-MT-10-072, Extended Power Uprate: Updates to Docketed Information, L-MT-11-044, Uprate (Epu): Update on EPU Commitments, L-MT-12-056, WCAP-17549-NP, Rev. 0, Monticello Replacement Steam Dryer Structural Evaluation for High-Cycle Acoustic Loads Using ACE, L-MT-12-090, Westinghouse, LTR-A&SA-12-8, Rev. 1, Attachment B, Recommendations for Inspections of the Monticello Replacement Steam Dryer., L-MT-12-114, Drawing C.5-2007, Rev. 17, Failure to Scram, L-MT-13-020, Enclosure 1 - Responses to the Gap Analysis, L-MT-13-029, Enclosure 14 to L-MT-13-029 - WCAP-17716-NP, Revision 0, Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project., L-MT-13-091, WCAP-17716-NP, Revision 1 - Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project, Enclosure 14, L-MT-13-092, Extended Power Uprate (Epu): Completion of EPU Commitments, Proposed License Conditions and Revised Power Ascension Test Plan, L-MT-13-096, GE-MNGP-AEP-3304R1, Enclosure 4, NEDC-33435 Corrected Pages, L-MT-13-126, Enclosure 3, GE-MNGP-AEP-3306, Enclosure 2 GEH Response to Mella+ Eicb RAI and Affidavit, L-MT-14-023, Maximum Extended Load Line Limit Analysis Plus: Technical Specification Clarifications, L-MT-15-074, Enclosure 7, WCAP-18604-NP, Revision 0, Monticello EPU Main Steam Line Strain Data Evaluation Report., L-MT-16-017, Revised Commitment to Reconcile Analysis of Bypass Voiding for Transition to Areva Analysis Methodology, L-MT-16-071, Submittal of 2016 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46, ML083230111, ML083400402, ML083500575, ML083570610, ML083590127, ML090710680, ML090710683, ML090710684, ML091120578, ML091140470, ML091410121, ML091410122, ML091410123, ML091410124, ML091760769, ML092290250, ML092810554, ML093160816, ML093220925, ML093220964, ML093620024, ML100280558... further results
MONTHYEAR2009-02-24 [Table View]

ML100280558