L-MT-14-046, Submittal of the Pressure and Temperature Limits Report (Ptlr), Revision 1

Submittal of the Pressure and Temperature Limits Report (Ptlr), Revision 1

ML14246A206
Person / Time
Site:	Monticello
Issue date:	08/28/2014
From:	Fili K Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-14-046
Download: ML14246A206 (42)

Text

Monticello Nuclear Generating Plant XcelEnergy 2807 W County Road 75 Monticello, MN 55362 August 28, 2014 L-MT-14-076 Technical Specification 5.6.5 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Submittal of the Pressure and Temperature Limits Report (PTLR), Revision 1 Northern States Power Company - Minnesota (NSPM), a Minnesota corporation, doing business as Xcel Energy, is providing in accordance with Technical Specification 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"a revised PTLR for the Monticello Nuclear Generating Plant. The PTLR provides the RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing established applying U.S. Nuclear Regulatory Commission approved methodologies. The current PTLR has been updated to reflect operation under a vacuum in the reactor pressure vessel.

Summary of Commitments This letter proposes no new commitments and does not revise any existing commitments.

Should you have questions regarding this letter, please contact Mr. Richard Loeffler at (763) 295-1247.

Executed on AugustA

, 2014.

Karen D. Fili Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region III, USNRC Resident Inspector, Monticello, USNRC Project Manager, Monticello, USNRC Minnesota Department of Commerce

ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

UP TO 54 EFFECTIVE FULL-POWER YEARS (EFPY)

REVISION 1 (39 pages follow)

Monticello Nuclear Generating Plant PTLR Revision 1 Page 1 of 39 SXceIEnergy@

Monticello Nuclear Generating Plant Pressure and Temperature Limits Report (PTLR) up to 54 Effective Full-Power Years (EFPY)

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Monticello Nuclear Generating Plant PTLR Revision I Page 2 of 39 Table of Contents Section Page Revision Record 3

1.0 Purpose 4

2.0 Applicability 4

3.0 Methodology 5

4.0 Operating Limits 6

5.0 Discussion 7

6.0 Plant Specific Information 12 7.0 References 16 Figure 1 MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 18 for 36 EFPY Figure 2 MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 19 for 40 EFPY Figure 3 MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 20 for 54 EFPY Figure 4 MNGP P-T Curve B (Normal Operation - Core Not Critical) 21 for 54 EFPY Figure 5 MNGP P-T Curve C (Normal Operation - Core Critical) 22 for 54 EFPY Table 1 MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY 23 Table 2 MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY 26 Table 3 MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY 29 Table 4 MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY 32 Table 5 MNGP Core Critical (Curve C) P-T Curves for 54 EFPY 35 Table 6 MNGP ART Calculations for 36 EFPY 36 Table 7 MNGP ART Calculations for 40 EFPY 37 Table 8 MNGP ART Calculations for 54 EFPY 38 Appendix A Monticello Reactor Vessel Materials Surveillance Program 39

Monticello Nuclear Generating Plant PTLR Revision 1 Page 3 of 39 REVISION RECORD Revision Description Initial Issue Revision to address operation of the RPV under a vacuum.

Revision No.

0

Monticello Nuclear Generating Plant PTLR Revision I Page 4 of 39 1.0 Purpose The purpose of the Monticello Nuclear Generating Plant (MNGP) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

I. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class I Leak Testing;

2. RCS Heatup and Cooldown rates;

3. Reactor Pressure Vessel (RPV) to RCS coolant AT (ATemperature) requirements during Recirculation Pump startups;

4. RPV bottom head coolant temperature to RPV coolant temperature AT requirements during Recirculation Pump startups;

5. RPV boltup temperature limits.

This report has been prepared in accordance with the requirements of Reference [1], Licensing Topical Report SIR-05-044-A, Revision 0, April 2007.

2.0 Applicability This report is applicable to the MNGP RPV up to 54 Effective Full-Power Years (EFPY).

The following MNGP Technical Specification (TS) is affected by the information contained in this report:

TS 3.4.9 RCS Pressure and Temperature (P-T) Limits

Monticello Nuclear Generating Plant PTLR Revision 1 Page 5 of 39 3.0 Methodology The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1], which has been approved for BWR use by the NRC.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [2], as documented in Reference [3].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 [4], as documented in Reference [5].

4. The pressure and temperature limits were calculated in accordance with Reference [1],

"Pressure - Temperature Limits Report Methodology for Boiling Water Reactors," April 2007, as documented in Reference [6].

5. This revision of the pressure and temperature limits is to incorporate the following changes:

0 Revision 1: Operation of the RPV under a vacuum.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Monticello Nuclear Generating Plant PTLR Revision I Page 6 of 39 Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV, cannot be made without prior NRC approval.

Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 54 EFPY for Monticello Nuclear Generating Plant, as documented in Reference [6]. The minimum required leak test temperature (Curve A) at 54 EFPY is above 2007F.

Because of the operational challenges presented by this elevated temperature, additional Curve A limits were developed at intermediate levels of 36 and 40 EFPY. Curve B and Curve C limits were not developed at 36 and 40 EFPY because the 54 EFPY limits for these curves do not present an operational challenge to MNGP. The MNGP Curve A limits for 36 EFPY are provided in Figure 1, and a tabulation of the curves is included in Table 1. The MNGP Curve A limits for 40 EFPY are provided in Figure 2, and a tabulation of the curves is included in Table 2. The MNGP P-T curves for 54 EFPY are provided in Figures 3 through 5, and a tabulation of the curves is included in Tables 3 through 5. The adjusted reference temperature (ART) tables for the MNGP vessel beltline materials are shown in Table 6 for 36 EFPY, Table 7 for 40 EFPY, and Table 8 for 54 EFPY (Reference [5]). The resulting P-T curves are based on the geometry, design and materials information for the MNGP vessel. The following conditions apply to operation of the MNGP vessel:

Monticello Nuclear Generating Plant PTLR Revision 1 Page 7 of 39

" Heatup and Cooldown rate limit during Hydrostatic Class 1 Leak Testing (Figures 1 through 3: Curve A): *_ 25°F/hour, [1].

• Normal Operating Heatup and Cooldown rate limit (Figure 4: Curve B - core non-critical, and Figure 5: Curve C - core critical): < 100°F/hour2 [6].

• Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 50'F.

RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 145°F.

RPV head flange, RPV flange and adjacent shell temperature limit during vessel bolt-up

_> 60-F [6].

5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG 1.99) [4] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the MNGP vessel plate, weld, and forging materials [5]; this evaluation included the results of three surveillance capsules. The Cu and Ni values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings.

Interpreted as the temperature change in any I-hour period is less than or equal to 251F.

2 Interpreted as the temperature change in any I -hour period is less than or equal to I 00TF.

Monticello Nuclear Generating Plant PTLR Revision 1 Page 8 of 39 The peak RPV ID fluence value of 6.43 x 1018 n/cm2 at 54 EFPY used in the P-T curve evaluation was obtained from Reference [3] and is calculated in accordance with RG 1.190 [2].

The intermediate peak RPV ID fluence values of 2.77 x 1018 n/cm2 at 36 EFPY and 3.36 x 1018 n/cm2 at 40 EFPY are calculated in [5] based on the flux values in [3]. The flux values in [3] are calculated in accordance with RG 1.190. Calculation details for intermediate fluence values, including benchmarking to the peak RPV ID fluence at 54 EFPY in [3], are given in [5, Appendix A]. These fluence values apply to the limiting beltline lower intermediate shell plates (Heat No. C2220-1 and C2220-2). The fluence values for the lower intermediate shell plates are based upon an attenuation factor of 0.738 for a postulated l/4T flaw. As a result, the l/4T fluence for the limiting lower intermediate shell plate is 2.04 x 1018 n/cm 2 at 36 EFPY, 2.48 x 1018 n/cm2 at 40 EFPY, and 4.75 x 1018 n/cm2 at 54 EFPY for MNGP.

The RPV ID fluence value of 1.01 x 10" n/cm2 at 54 EFPY used in the P-T curve evaluation of the recirculation inlet nozzle was obtained from Reference [5] and is calculated in accordance with RG 1.190 [2]. The intermediate RPV ID fluence values of 4.27 x 1017 n/cm 2 at 36 EFPY and 5.23 x 1017 n/cm 2 at 40 EFPY are calculated in [5] based on the flux values in [3]. The flux values in [3] are calculated in accordance with RG 1.190. Calculation details for intermediate fluence values, including benchmarking to the peak RPV ID fluence at 54 EFPY in [3], are given in [5, Appendix A]. These fluence values apply to the limiting recirculation inlet nozzle (Heat No. E21VW). The fluence value for the recirculation inlet nozzle is based upon an attenuation factor of 0.738 for a postulated l/4T flaw.

As a result, the 1/4T fluence for the limiting recirculation inlet nozzle is 3.151 x 1017 n/cm 2 at 36 EFPY, 3.86 x 1017 n/cm 2 at 40 EFPY, and 7.45 x 1017 n/cm 2 at 54 EFPY for MNGP. There are no additional forged or instrument nozzles in the extended beltline at 54 EFPY.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T

Monticello Nuclear Generating Plant PTLR Revision I Page 9 of 39 location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup.

However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature.

This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, which is well above the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of < 1 00°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound RPV thermal transients. For the hydrostatic pressure and leak test curves (Curve A), a coolant heatup and cooldown temperature rate of_< 25°F/hr must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RTNDT, the chemistry (weight-percent copper and nickel) and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 101 7 n/cm 2 for E > 1MeV) are shown in Table 6 for 36 EFPY, Table 7 for 40 EFPY, and Table 8 for 54 EFPY [5].

Per Reference [5] and in accordance with Appendix A of Reference [1], the MNGP representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). The representative heat of the plate material (C2220) in the ISP is the same as the lower intermediate

Monticello Nuclear Generating Plant PTLR Revision I Page 10 of 39 shell plate material in the vessel beltline region of MNGP. For plate heat C2220, since the scatter in the fitted results is less than I-sigma (I7°F), the margin term (aA = 17°F) is cut in half for the plate material when calculating the ART. The representative heat of the weld material (5P6756) in the ISP is not the same as the limiting weld material in the vessel beltline region of MNGP.

Therefore, CFs from the tables in RG 1.99 were used in the determination of the ART values for all MNGP materials except for plate heat C2220.

The only computer code used in the determination of the MNGP P-T curves was the ANSYS Mechanical and PrepPost, Release 11.0 (with Service Pack 1) [7] finite element computer program for the feedwater nozzle (non-beltline) and recirculation inlet nozzle (beltline) stresses.

The ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B [8] Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [9] was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific MNGP feedwater nozzle analysis was performed to determine through-wall thermal and pressure stress distributions due to a bounding thermal transient [10]. Detailed information regarding the analysis can be found in Reference [10]. The following inputs were used as input to the finite element analysis:

With respect to operating conditions, stress distributions were developed for two bounding thermal transients. A thermal shock, which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions, and a thermal ramp were analyzed [10]. Because the feedwater nozzle thermal sleeve is an integral part of the safe-end, the thermal shock that occurs in the feedwater nozzle as part of the startup transient is significantly reduced. As a result, the thermal ramp of 100°F/hr, which is

Monticello Nuclear Generating Plant PTLR Revision I Page 11 of 39 associated with the shutdown transient, produces higher tensile stresses at the 1/4T location. Therefore, the stresses represent the bounding stresses in the feedwater nozzle associated with I 00*F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region.

" Heat transfer coefficients were given in the MNGP feedwater nozzle governing basis stress report for both forced and free convection in the vessel. The analysis used the higher forced convection coefficient of 500 Btu/hr-ft2-°F, and applied it to all wetted surfaces [10]. Therefore, the heat transfer coefficients used in the analysis bound the actual operating conditions in the feedwater nozzle at MNGP.

" A one-quarter symmetric, three-dimensional finite element model of the feedwater nozzle was constructed (Reference 10). Temperature dependent material properties, taken from the MNGP Code of Record [11], were used in the evaluation.

The plant-specific MVNGP recirculation inlet nozzle analysis was performed to determine through-wall thermal and pressure stress distributions due to a bounding thermal transient [12].

Detailed information regarding the analysis can be found in Reference [12].

The following inputs were used as input to the finite element analysis:

With respect to operating conditions, the thermal transient that would produce the highest tensile stresses at the 1/4T location is the 100°F/hr shutdown transient [12]. Therefore, the stresses represent the bounding stresses in the recirculation inlet nozzle associated with 100*F/hr heatup/cooldown limits associated with the P-T curves for a nozzle in the beltline region.

Heat transfer coefficients were calculated in accordance with the MINGP recirculation inlet nozzle governing basis stress report.

The heat transfer coefficients were conservatively based on the full temperature difference of the transient, rather than the RPV to coolant temperature difference [12].

The nozzle blend radius heat transfer

Monticello Nuclear Generating Plant PTLR Revision I Page 12 of 39 coefficient used the higher of the calculated vessel heat transfer coefficient (675 Btu/hr-ft-°F) or the calculated nozzle heat transfer coefficient (265 Btu/hr-ft2-°F). Therefore, the heat transfer coefficients used in the analysis bound the actual operating conditions in the recirculation inlet nozzle at MNGP.

A one-quarter symmetric, three-dimensional finite element model of the recirculation inlet nozzle was constructed (Reference 12). Temperature dependent material properties, taken from the MNGP Code of Record [11], were used in the evaluation.

Reference [13] contains NRC approval of Monticello initial RTNDT values which are used in development of the Pressure-Temperature limits documented in this PTLR.

6.0 Plant-Specific Information EPU vs. MELLLA+ Fluence Calculations MNGP is planning to implement both an extended power uprate (EPU) to 2004 MWth and MELLLA+ (Maximum Extended Load Line Limit Analysis) operation during the current operating license. In preparation for these changes, fluence calculations were performed in accordance with Reg Guide 1.190 to determine the effects on the flux profile of the reactor vessel and its internals. In 2007, a fluence calculation was developed to determine the projected fluence accumulation for the reactor vessel considering EPU power level (2004 MWTH) to the end of the current operating license (54 EFPY/2030). In 2009, an additional fluence calculation was performed to consider EPU power levels with MELLLA+ operation to the end of the current operating license at 2004 MWth. Since both calculations were developed in accordance with Reg Guide 1.190, the fluence values for components used to determine adjusted reference temperature and related pressure-temperature limits and curves were compared in the fluence calculations and the most conservative value was used. For all components, the 2007 EPU-only fluence calculation was more conservative than the EPU/MELLLA+ values and the EPU-only values were used in the determination of the pressure-temperature limits and curves in the PTLR.

Monticello Nuclear Generating Plant PTLR Revision 1 Page 13 of 39 Excess Conservatism in Fluence and Multiple Curves for Hydrostatic Pressure Test While reviewing the EPU-only fluence calculation, it was determined that fluence values for locations with an accumulated fluence nearer to the lower bound of 1.0 x 1017 n/cm 2 were overly conservative. The fluence values for these locations were given an additional factor of 1.3 to account for potential variation in future operation and assumed EPU was implemented after Cycle 22 in 2005 (28.82 EFPY). The overly conservative fluence values resulted in hydrostatic pressure test temperatures near 2127F. With pressure test temperatures near 212'F, additional preparations must be made in case of entry into Mode 3 during the pressure test. These additional preparations will result in longer outage durations, additional dose and more risk to the site and site personnel.

In order to avoid entry in Mode 3, some of the conservatism was removed from the fluence values for the upper intermediate shell plates, lower shell plates and the N-2 Nozzles. The conservatism was removed by applying the 1.3 factor only to operation past EPU (33.4 EFPY) for fluence calculated at 36 EFPY and 40 EFPY. Even with the excess conservatism removed, the fluence values are conservative because a review of past operation and fluence accumulation on the reactor vessel show that conditions before 33.4 EFPY (2011) are bounded by pre-EPU fluence without the 1.3 factor. 33.4 EFPY is determined to be the EFPY as of April 2011 and for the purposes of this evaluation and to maintain margin and conservatism it is assumed to be the beginning of EPU implementation. The flux values used to calculate the fluence values with the excess conservatism removed were calculated in accordance with NRC Reg Guide 1.190. The fluence values with the removed excess conservatism were calculated at 36, 40 and 54 EFPY are shown in the following table.

Monticello Nuclear Generating Plant PTLR Revision 1 Page 14 of 39 Component Fluence[5]

RPV Component 36 EFPY 40 EFPY 54 EFPY 77/c6) 2 ni/cm2 fI/C))2 Upper Intermediate Shell Plates 1.97x10 7 2.30x10l7 4.06x10 7 (1-12 and 1-13)

Lower Intermediate Shell Plates 2.77x101 3.36x018 6.43x10 8

(-14 and 1-15)

Lower Shell Plates (1-16 and 1-17) 1.85x10'8 2.28x10's 4.46x10'8 Limiting Weld 2.77x1018 3.36x10'8 6.43x10'8 N-2 Nozzles 4.27x1017 5.23x10 17 1.01xlIO1 Each of the various curves will be used for the hydrostatic pressure-test required at the end of each refueling outage. The curve that will be used for a specific outage will be determined by the accumulated fluence on the vessel. The hydrostatic test procedure will include a step to verify the vessel accumulation and determine which curve will bound the current vessel fluence accumulation for use in that specific outage.

Monticello 3000 Surveillance Capsule In 2007, Monticello sent the surveillance capsule located at the 3000 reactor vessel azimuth out for testing in accordance with the requirements of the BWRVIP Integrated Surveillance Program (ISP) of which Monticello is an active member. The results of the testing were received in March 2009 and in accordance with the requirement of the ISP and Reg Guide 1.190, these results must be included in fluence calculations for the development of any pressure-temperature limits including the PTLR. Since the fluence calculation used for developing the pressure-temperature limits was completed in 2007, the surveillance capsule results were not included in the fluence evaluation. In order to incorporate the 2009 surveillance capsule data, the results were evaluated by General Electric to determine if the fluence accumulated by the Surveillance capsule was

Monticello Nuclear Generating Plant PTLR Revision I Page 15 of 39 within the uncertainty range of the fluence calculation performed in 2007. GE found that the fluence capsule data was within the uncertainty range of the fluence calculation from 2007. [15]

Operation of the RPV Under a Vacuum In April 2014, it was discovered that the RPV is operated under a vacuum during startup operations. Drawing a vacuum on the RPV was evaluated and it was determined that the RPV remains acceptable at all conditions including a slight vacuum caused by the pre-start up evolution.

During startup, operations closes the RPV Head Vent lines, opens the Main Steam Isolation Valves (MSIVs) to the condenser and places the Mechanical Vacuum Pump (MVP) in service.

This configuration allows a slight vacuum to be drawn on the upper portion of the vessel. The straight line on the hydrostatic (Curve A), core not critical (Curve B), and core critical (Curve C) curves terminates at 0 psig. The straight line on Curves A, B and C is representative of the 10CFR50 Appendix G minimum required temperature for the RPV in low pressure conditions.

The PTLR curves have traditionally started at 0 psig for simplification of the curve illustration and based on pressure sensor equipment limitations. However, I OCFR50 Appendix G has no lower pressure limitation related to the minimum required temperature. The 1 OCFR50 Appendix G requirement for low vessel pressures is to maintain the minimum metal temperature during both normal operation and hydrostatic pressure and leak tests. This requirement also includes vessel pressures below 0 psig. Based on the discovery, a note was added to Curves A, B and C to indicate that the P-T limit at 0 psig is applicable for RPV operation under vacuum as long as the minimum required vessel temperature is maintained in accordance with 10CFR50 Appendix G.

Monticello Nuclear Generating Plant PTLR Revision I Page 16 of 39 7.0 References

1. Structural Integrity Associates Report No.

SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007, SI FileNo. GE-10Q-401.

2. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and DosimetryMethods for Determining Pressure Vessel Neutron Fluence", March 2001.

3. MNGP Site Calculation 11-039, "Monticello Neutron Flux and Fluence Evaluation for Extended Power Uprate," December 2007

4. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel materials," May 1988.

5. Structural Integrity Associates Calculation No. 1000847.301, Revision 2, "Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts," July 2011.

(MNGP Site Calculation 11-003, Rev. OA)

6. Structural Integrity Associates Calculation No. 1000847.303, Revision 2, "Revised P-T Curves Calculation," August 2011. (MNGP Site Calculation 11-005, Rev. 0A)

7. ANSYS Mechanical and PrepPost, Release 11.0 (w/Service Pack 1), ANSYS, Inc.,

August 2007.

8. U. S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".

9. U. S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, "License Qualification for Performing Safety Analyses," June 24, 1999.

10. Structural Integrity Associates Calculation No.

1000847.302, Revision 0, "Finite Element Stress Analysis of Monticello RPV Feedwater Nozzle," October 2010. (MNGP Site Calculation 11-004, Rev. 0)

Monticello Nuclear Generating Plant PTLR Revision I Page 17 of 39

11. ASME Boiler and Pressure Vessel Code,Section III including Appendices, 1977 Edition with Addenda through Summer 1978.

12. Structural Integrity Associates Calculation No. 1000720.301, Revision 0, "Finite Element Stress Analysis of Monticello RPV Recirculation Inlet Nozzle," June 2010. (MNGP Calculation 11-020, Rev. 0)

13. NRC ( C.F. Lyon) letter to NMC (R.O Anderson), "Monticello Nuclear Generating Plant-Issuance of Amendment RE: Revision of Reactor Vessel Pressure-Temperature Limit Curves and Removal of Standby Liquid Control Relief Valve Setpoint (TAC No.

MA4532)", dated October 12, 1999.

14. NRC (L. M. Padovan) letter to NMC (D. L. Wilson), "Monticello Nuclear Generating Plant - Issuance of Amendment re: Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program (TAC No. MB6460)", dated April 22, 2003.

15. G.E Letter Number 0000-0122-7030, Revision 0, "Calculation-to-Measurement Ratio of Monticello 300-Degree Surveillance Capsule",

August

2010, CONTAINS PROPRIETARY INFORMATION

16. EC 23962, Revision 0, "Structural Integrity of the Reactor Vessel (RPV) under a Vacuum", April 2014

17. EC 23963, Revision 0, "Monticello P-T Curve Limit Violation Assessment Vendor-Prepared", April 2014

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E40*

COMPUANCE REQUIRES OPERATION ABOVETHECURVES j

.1 1

1 1 1.1 I

111111 1111111111111111! itm M! 111111 111111 11 1 1 1 1! 111111 11 cii.

9 z

3 9

3 0

9 3

3 0

0 9

z 0

0 9

0 9

(A I

][1F1[]1[1tH~'I]t1i1[1t1[1fl[1"L~4

..........- _. I......

I I I I I 1 1.-4 1-1 T I I Beitfine Region

-Bottom Head tBoft-up Temp:

M___L1_

A I UpperVessel 60*F I

I I I I I I I I I I I I I I I I T

1111111 20 0

00 O

0 0

0 100 200 300 400 500 600 700 g0 S0 1000 1100 1200 1300 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (pslg)

Note: The minimum reactor vessel metal temperature at 0 psig is applicable for RPV operation under a vacuum.

300 280 260 240 U-D 220 200 180 160

-J 140 CO) w 120 S100 80 60 40 20 E _1_0ýý COMPUANCE REQUIRES OPERA71ON A VETHqECURVES I I I 1 1-1 v 1. 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 U -U 1 11411

-I U'

(-I 0

• 1 0

• 1 0

0

-I 0

-I U'

i i i i*i i ii i i i iii

'iii i i l l l i l l ~ l i i I

Minimum Criticality Temp:

70 *F ILL I I I I I LL I I I I I 1.1-ILLI.J. I J-1. I M 0

0 z

CD CD CD cJQ 0"

w~0 0

0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 PRESSURE LIMIT IN REACTOR VESSEL TOP HEAD (psig)

Note: The minimum reactor vessel metal temperature at 0 psig is applicable for RPV operation under a vacuum.

Monticello Nuclear Generating Plant PTLR Revision I Page 23 of 39 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY Beltline Region P-T Curve Temperature 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 61.75 86.10 102.40 114.67 124.52 132.74 139.81 146.00 151.49 156.45 160.96 165.09 168.91 172.47 175.78 178.88 181.82 184.58 187.19 189.69 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Monticello Nuclear Generating Plant PTLR Revision I Page 24 of 39 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R =

ART =

Kit =

Safety Factor =

Stress Concentration Factor =

Mm =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment =

(penetrations portion) inches inches

-F (no thermal effects)

(bottom head penetrations)

"F (applied after bolt-up, instrument uncertainty) inches psig (hydrostatic pressure head for a full vessel at 70"F) psig (instrument uncertainty)

Gauge Fluid Temperature I*F)

Temperature KIC Kim for P-T Curve (ksl*inch1I2)

(ksi*inch112)

(OF)

Adjusted Pressure for P-T Curve (Dsia)

I I

60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 74.13 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 49.42 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 0

813 832 851 872 893 915 939 963 988 1,014 1,041 1,069 1,099 1,129 1,161 1,194 1,229 1,265 1,302

Monticello Nuclear Generating Plant PTLR Revision 1 Page 25 of 39 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY (continued)

Plant =

Component =

ART:

Vessel Radius, R =

Nozzle comer thickness. t' =

Kit Crack Depth, a =

Safety Factor =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment Reference Pressure Unit Pressure Flange RTNOT =

Inches Inches. approximate (no thermal effects) ksi*inch'r2 inches

'F (applied after bolt-up, instrument uncertainty) inches psig (hydrostatic pressure head for a full vessel at 70"F) pslg (instrument uncertainty) psig (pressure at which the FEA stress coefficients are %slid) pslg (hydrostatic pressure)

"F==-==.>

All EFPY Gauge Fluid Temperature

(*F)

P-T Curve Temperature

• F)

P-T Curve IOCFR60 Adjustments (psug)

(ksr~lnch11t )

(ksl~lnch115 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 108.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 122.0 124.0 64.13 64.13 65.39 66.71 68.08 69.50 70.98 72.52 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.58 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 120.71 124.28 128.00 131.87 135.90 140.09 144.45 42.75 42.75 43.60 44.47 45.38 46.33 47.32 48.35 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 80.47 82.86 85.33 87.91 90.60 93.39 90.30 60 60 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 102 104 106 108 110 112 114 116 118 120 122 124 0

313 313 616 629 643 657 672 688 704 721 738 756 775 795 815 837 859 882 908 931 957 984 1012 1042 1072 1104 1137 1172 1207 1245 1284 1324 1366

Monticello Nuclear Generating Plant PTLR Revision I Page 26 of 39 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY Beltline Region P-T Curve Temperature 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 67.65 92.00 108.30 120.57 130.42 138.64 145.71 151.89 157.39 162.35 166.86 170.99 174.81 178.37 181.68 184.78 187.71 190.48 193.41 196.73 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Monticello Nuclear Generating Plant PTLR Revision I Page 27 of 39 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R =

ART =

Kit =

Safety Factor =

Stress Concentration Factor =

Mm =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment =

Gauge Fluid Temperature (OF)

BottomiIHead (penetrations portion) inches inches 2.0 *F 0.00 "(no thermal effects)

(ottom head penetrations) 0,

• F (applied after bolt-up, instrument uncertainty) 758.00 inches 27ý4 psig (hydrostatic pressure head for a full vessel at 70*F) 0 psig (instrument uncertainty)

Adjusted Temperature Pressure for KIcK m

for P-T Curve P-T Curve Iksl*lnchlI 2I (ksi*lnchl/ 21 (OF)

(Dsl) 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 74.13 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 49.42 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 0

813 832 851 872 893 915 939 963 988 1,014 1,041 1,069 1,099 1,129 1,161 1,194 1,229 1,265 1,302

Monticello Nuclear Generating Plant PTLR Revision 1 Page 28 of 39 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY (continued)

Plant =

Component =

ART Vessel Radius, R =

Nozzle comer thickness, t' =

Kip-appwi.

Crack Depth, a =

Safety Factor =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Acdustment =

Reference Pressure =

Unit Pressure =

Flange RTNoT =

Inches Inches, approximate (no thermal effects) ksl*lnchl" inches "F (applied after bolt-up, instrument uncertainty)

Inches psig (hydrostatic pressure head for a full vessel at 70'F) psig (Instrument uLcertalnty) psig (pressure at which the FEA stress coefficients are %elid) pslg (hydrostatic pressure)

°F======>

All EFPY Gauge Fluid Temperature KC Iqp

• F1 (ksi*lnch"21 (ksl*inch212)

P-T P-T Curve Curve 10CFRSO Temperature Adjustments

('F1 (folal 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 122.0 124.0 64.13 64.13 65.39 66.71 68.08 69.50 70.98 72.52 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.68 91.86 94.25 96.76 99.34 102.04 104.85 107.77 110.82 113.98 117.28 120.71 124.28 128.00 131.87 135.90 140.09 144.45 42.75 42.75 43.60 44.47 45.38 46.33 47.32 48.35 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 80.47 82.86 85.33 87.91 90.60 93.39 96.30 60 60 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 102 104 108 108 110 112 114 116 118 120 122 124 0

313 313 616 629 643 657 672 688 704 721 738 756 775 795 815 837 859 882 908 931 957 984 1012 1042 1072 1104 1137 1172 1207 1245 1264 1324 1366

Monticello Nuclear Generating Plant PTLR Revision 1 Page 29 of 39 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY Beitline Region P-T Curve Temperature 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 82.85 107.19 123.50 135.78 145.62 153.85 160.90 167.09 172.59 177.55 184.05 191.16 197.39 202.93 207.92 212.45 216.61 220.44 224.02 227.33 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Monticello Nuclear Generating Plant PTLR Revision I Page 30 of 39 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R =

ART =

Kit =

Safety Factor =

Stress Concentration Factor =

Mm =

Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment =

OF (applied after bolt-up, instrument uncertainty) inches psig (hydrostatic pressure head for a full vessel at 70=F) psig (instrument uncertainty)

Gauge Fluid Temperature

(*F)

Temperature KC KIm for P-T Curve (ksl*lnchlI 2)

(ksi*inchlI 2)

(OF)

Adjusted Pressure for P-T Curve (psig) 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 74.13 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 49.42 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 0

813 832 851 872 893 915 939 963 988 1,014 1,041 1,069 1,099 1,129 1,161 1,194 1,229 1,265 1,302

Monticello Nuclear Generating Plant PTLR Revision 1 Page 31 of 39 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Plant -'

l$9P Component lTIpeaso ART 40.0:

Vessel Radius, R =

Nozzle corner thickness, tr =

7 Kipt =I Crack Depth, a 1.33 Safety Factor- =

.o Temperature Adjustment =I' Height of Water for a Full Vessel =

Pressure Adjustment =

27 Pressure Adjustment =

0 Reference Pressure 1,000 Unit Pressure -

663 Flange RTNoT =

Gauge Fluid Temperature KNI

("F)

(ksl*inchl2) nches nches, approximate no thermal effects) csl*incht'2 nches

'F (applied after bolt-up, Instrument uncertainty) nches psig (hydrostatic pressure head for a full %essel at 70*F) psig (instrument uncertainty) pslg (pressure at which the FEA stress coefficients are %slid) sig (hydrostatic pressure)

°F......

All EFPY P-T Curve K14 Temperature fksl*lnch1 21 PFI P-T Curve IOCFR5O Adjustments (oslal "1

t

,r-60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 780 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 122.0 124.0 64.13 64.13 65.39 66.71 68.08 69.50 70.98 72.52 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 120.71 124.28 128.00 131.87 135.90 140.09 144.45 42.75 42.75 43.60 44.47 45.38 46.33 47.32 48.35 49.42 50.53 51.69 52.90 54.15 55.46 56.82 58.23 59.71 61.24 62.84 64.50 66.23 68.03 69.90 71.85 73.88 75.99 78.19 80.47 82.86 85.33 87.91 90.60 93.39 96.30 60 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 102 104 106 108 110 112 114 116 118 120 122 124 0

313 313 616 629 643 857 672 688 704 721 738 756 775 795 815 837 859 882 906 931 957 984 1012 1042 1072 1104 1137 1172 1207 1245 1284 1324 1366

Monticello Nuclear Generating Plant PTLR Revision I Page 32 of 39 Table 4: MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY Beltline Region P-T Curve Temperature 60.00 60.00 60.00 60.00 89.07 116.72 134.43 137.89 138.16 147.47 157.81 166.37 174.76 185.75 194.74 202.37 208.98 214.82 220.05 224.78 229.10 233.08 236.77 240.21 243.41 246.43 249.28 251.98 254.53 P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Monticello Nuclear Generating Plant PTLR Revision 1 Page 33 of 39 Table 4: MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Plant =

Component =

Bottom Head thickness, t =

Bottom Head Radius, R -

ART=

Kit =

Safety Factor =

Stress Concentration Factor =

Mm:

Temperature Adjustment Height of Water for a Full Vessel =

Pressure Adjustment =

Pressure Adjustment =

Heat Up and Cool Down Rate = I (penetrations portion)

'inches inches

-°F kslinch t1 2

(bottom head penetrations)

°F (applied after bolt-up, Instrument uncertainty)

Inches psig (hydrostatic pressure head for a full vessel at 70°F) psig (instrument uncertainty)

°F/Hr Gauge Fluid Temperature

(*F)

Temperature Kic Kim for P-T Curve (ksl*inchll)

(ksi*lnchuIS)

(F)

Adjusted Pressure for P-T Curve (psig) 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 74.13 74.13 75.80 77.54 79.34 81.23 83.19 85.23 87.35 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 120.71 124.28 128.00 131.87 135.90 140.09 144.45 148.99 153.72 158.63 163.75 169.08 32.97 32.97 33.81 34.67 35.58 36.52 37.50 38.52 39.58 40.69 41.84 43.03 44.28 45.58 46.93 48.33 49.79 51.31 52.90 54.55 56.26 58.05 59.91 61.84 63.85 65.95 68.13 70.40 72.76 75.22 77.78 80.45 60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 0

533 547 562 578 594 610 628 646 664 684 704 725 747 770 794 819 845 872 900 929 960 991 1,024 1,058 1,094 1,131 1,170 1,210 1,251 1,295 1,340

Monticello Nuclear Generating Plant PTLR Revision I Page 34 of 39 Table 4: MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Plant Component.

ART:

Vessel Radius. R Nozzle comer thickness, tI K4 1' Crack Depth. a.

Safety Factor =

Temperature Adjustment=

Height of Water for a Full Vessel Pressure Adjustment.

Pressure Adjustment.

Reference Pressure Unit Pressure' Flange RTNoT rnches riches, approxkimate tsl*lnchWI riches F (applied after boll-up, Instnruent uncertainty) riches slg (hydrostatic pressure head for a full %assel at 70*F) wig (Instrument uncertakity) sig (pressure at which the FEA stress coefficients are wild) psig (hydrostatic pressure)

F=====.>

All EFPY Gauge P-T P-T Fluid Curve Curve Temperature K,.

Ke Temperature Pressure

(*F)

(ksli*nch 10 )

(kslnchwhm)

(*F)

(pWlg 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.0 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 122.0 124.0 126.0 128.0 130.0 132.0 134.0 136.0 138.0 140.0 142.0 64.13 28.53 64.13 28.53 65.39 29.16 66.71 29.82 68.08 30.51 69.50 31.22 70.98 31.96 72.52 32.73 74.13 33.53 75.80 34.37 77.54 35.24 79.34 36.14 81.23 37.08 83.19 38.06 85.23 39.08 87.35 40.14 89.56 41.25 91.86 42.40 94.25 43.60 96.75 44.84 99.34 4614 102.04 47.49 104.85 48.89 107.77 50.35 110.82 51.88 113.98 63.46 117.28 55.11 120.71 56.82 124.28 58.61 128.00 60.47 131.87 62.40 135.90 64.42 140.09 66.51 144.45 68.69 148.99 70.96 153.72 73.33 158.63 75.78 163.75 78.34 169.08 81.01 174.63 83.78 180.40 86.67 186.40 89.67 192.66 92.80 60 60 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 130 132 134 136 138 140 142 0

313 313 404 414 424 435 446 458 470 483 496 5Ow 523 538 554 570 586 603 622 640 660 680 701 723 746 770 795 821 848 876 905 935 967 1000 1034 1069 1106 1145 1185 1227 1270 1315

Monticello Nuclear Generating Plant PTLR Revision I Page 35 of 39 Table 5: MNGP Core Critical (Curve C) P-T Curves for 54 EFPY Plant Curve A Leak Test Temperature =

Curve A Pressure =

Unit Pressure =

Flange RTNDT =

P-T Curve Temperature 70.00 70.00 70.00 70.00 129.07 156.72 174.43 177.89 206.00 206.00 206.00 206.37 214.76 225.75 234.74 242.37 248.98 254.82 260.05 264.78 269.10 273.08 276.77 280.21 283.41 286.43 289.28 291.98 294.53 ipsig psig (hydrostatic pressure)

'OF P-T Curve Pressure 0

50 100 150 200 250 300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300

Monticello Nuclear Generating Plant PTLR Revision 1 Page 36 of 39 Table 6: MNGP ART Calculations for 36 EFPY j

I

" 9 1 Upper/nt Shel 1-13 5.063 1.266 1.97E+17 0.738 1.454E+17 0.141 Lower/Irt Shel 1-14 5.063 1.266 2.77E+18 0.738 2.044E+18 0.575 Lower/kit Shel 1-15 5.063 1.266 2.77E+1 8 0.738 2.044E+18 0.575 Lower Shel 1-16 5.063 1.266 1.85E+18 0.738 1.365E+18 0.482 Lower Shel 1-17 5.063 1.266 1.85E+18 0.738 1.365E+18 0.482 Limiting Weld - Belfne 5.063 1.266 2.77E+18 0.738 2.044E+18 0.575 Bounding N-2 Nozzle 5.063 1.266 4.27E+17 0.738 3.151E+I 7

0.226

Monticello Nuclear Generating Plant PTLR Revision 1 Page 37 of 39 Table 7: MNGP ART Calculations for 40 EFPY UpperInt Shell -12 C2089-1 0.0 0.35 0.50 199.50 31.0 15.5 0.0 61.9 Shel.13 02613-1 i

27.0 0.35 0.49 19825 30.8 15.4 0.0 88.6 Lower/int Shell 1-14 C2220-1 27.0 0.16 0.64 180.00 112.0 8.5 0.0 166.0 Lowerlint Shel 1-15 C2220-2 27.0 0.16 0.84 180.00 112.0 8.5

.0 Lower Shell 1-16 A0946-1 27.0 0.14 0.56 98.20 51.9 17.0 0.0

.9 Lower Shell 1-17 C2193-1 0.0 017 050 118.50 62.7 17.0 0.0 96.7 Lounding Weld Nozflne

-8ON45.60.0 09 3.0 839

2.

27 7.

Lmiting WheldlBe-1n 5.063 1.266 3.36E+18 0.738 2.48E+18 0.622 BnDing Wed-Betio e 5.063 1.266 3.T3E+18 0.738 2.48E+18 0.624 Limiting Wed-2 Nozzie 5.08318N6 65.63E+110 0.99 149 83.96E+1701.72 79.

Monticello Nuclear Generating Plant PTLR Revision I Page 38 of 39 Table 8: MNGP ART Calculations for 54 EFPY chornisry~ RdjostmientFor 114t Descriotio'n Code No.

-Heatt No.

Flux Type & Lot-No ~lnitialRTjr T KF Factor ARTývj :MarrfTeifi,,

AR~T~

Upper/int Shell 1-12 C2089-1 0.0 0.35 0.50 199.50 43.8 17.0 0.0 77.8 U_

pper/Ikt Shell 1-13 C2613-1 27.0

.35 0.4 198.25 17.0 0.0 104.5 Lowerfint Shen 1-14 C2220-1 27.0 0.16 0.64 180.00 14.6' 8.5 0.0 186.

LowerfintShen-1M C2220-2 27.0 0.16 0.64 180.00 142.6

8.

1.6 Lower Shen 1-16 A0946-1 27.0 0.14 0.56 98.20 68.2 17.0 0.0 129.2 Lower Shell W7 C2193-1 0.0 0.17 0.50 118.50 1

116.3 es ptio oý'

e R

Limiting Weld - Beldine E8018N

.65.6 0.10 0.99 134.90 106.9 28.0 12.7 102.8 Bounding N-2 Nozzle E21VW P late W-16 1-17ý 40.0 01 0.6 141.90 512 7.

Upper/knt Shell 1-12 5.063 1.266 4.06E+17 0.738 2.996E+17 0.219 Upper/knt Shen 1-13 5.063 1.266 4.06E+1 7 0.738 2.996E+1 7 0219 Lower/knt Shen 1-14 5.063 1.266 6.43E+18 0.738 4.746E+1 8 0.792 Lower/kit Shell 1-15 5.063 1.266 6.43E+18 0.738 4.746E+18 0.792 Lower Shell 1-16 5.063 1.266 4.46E+18 0.738 3.292E+1 8 0.694 Lower Shell 1-17 5.063 1.266 4.46E+18 0.738 3.292E+1 8 0.694 Limiting Weld - Berline 5.063 1.266 6.43E+1 8 0.738 4.746E+1 8 0.792 Bounding N-2 Nozzle 5.063 1.266 1.01 E+1 8 0.738 7,454E+17 0.361

Monticello Nuclear Generating Plant PTLR Revision 1 Page 39 of 39 Appendix A MONTICELLO REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, a surveillance capsule was removed from the Monticello reactor vessel in 2007.

The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region.

MNGP has made a licensing commitment to replace the existing surveillance program with the BWRVIP ISP, and intends to use the ISP for MNGP during the period of extended operation.

The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC. Xcel Energy committed to use the ISP in place of its existing surveillance programs in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated April 22, 2003 [14]. The surveillance capsule removed in 2007 contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. MNGP continues to be a host plant under the ISP. One more Monticello capsule is scheduled to be removed and tested under the ISP in approximately 2022.

QF0212, Revision 5 (FP-SC-RSI-04)

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