ML12033A176

License Amendment Request Revise the Technical Specifications to Include a Pressure Temperature Limits Report Pressure and Temperature Limits Report, Enclosure 4

ML12033A176
Person / Time
Site:	Monticello
Issue date:	09/08/2011
From:	Mcgruder W Xcel Energy
To:	Office of Nuclear Reactor Regulation
References
L-MT-12-002
Download: ML12033A176 (39)

Text

ENCLOSURE4 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISE THE TECHNICAL SPECIFICATIONS TO INCLUDE A PRESSURE TEMPERATURE LIMITS REPORT PRESSURE AND TEMPERATURE LIMITS REPORT (38 pages follow)

Monticello Nuclear Generating Plant PTLR Revision 0 Page 1 of 38

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

Technical Lead: Date:

Wynter McGruder Technical Review: Date:

ýirn Bi mn-an Program Manager:

Gary Sherwood

Monticello Nuclear Generating Plant PTLR Revision 0 Page 2 of 38 Table of Contents Section Paae 1.0 Purpose 3 2.0 Applicability 3 3.0 Methodology 4 4.0 Operating Linits 5 5.0 Discussion 6 6.0 Plant Specific Information 11 7.0 References 15 Figure i MGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 17 for 36 EFPY Figure 2 MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 18 for 40 EFPY Figure 3 MNGP P-T Curve A (Hydrostatic Pressure and Leak Tests) 19 for 54 EFPY Figure 4 MNGP P-T Curve B (Noimal Operation - Core Not Critical) 20 for 54 EFPY Figure 5 MNGP P-T Curve C (Normal Operation - Core Critical) 21 for 54 EFPY Table I MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY .22 Table 2 MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY 25 Table 3 MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY 28 Table 4 MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY 31 Table 5 MNGP Core Critical (Curve C) P-T Curves f6if 54 EFPY 34 Table 6 MNGP ART Calculations for 36 EFPY 35 Table 7 MNGP ART Calculations for 40 EFPY 36 Table 8 MNGP ART Calculations for 54 EFPY 37 Appendix A Monticello Reactor Ves~sel Materials Surveillance Program3 38

Monticello Nuclear Generating Plant PTLR Revision 0 Page 3 of 38 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;

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

2. RCS Heatupand 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.

20 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 0 Page 4 of 38 3.0 Methodoloyv The limits in this report were derived as follows:

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

2. Theneutron fluence is calculated in!accordance with NRC Regulatory Guide .190 (RG 1.190) [2], as documented in Reference [3].

3. The adjusted ireference 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:

Initial issue of PTLR, 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 0 Page 5 of 38 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 200'F. 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 Revisioin 0 Page 6 of 38 b Heatup and Cooldown rate limit during Hydrostatic Class I Leak Testing (Figures 1 through 3: Curve A): < 25T/hour' [1].

o Normal Operating Heatup and Cooldown rate limit (Figure 4; Curve B -core non-critical, and Figure 5: Curve C - core critical): < 100OF/hour 2 [6].

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

" RPV bottom head coolant temperature toRPV coolant temperature AT limit during Recirculation Punp startup: < 145T.

o 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 kG 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 I of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for Welds. The Cu and Ni valueswere 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 1-hour period is less than or equal to 25*F.

2 Interpreted as the temperature change in any I-hour period is less than or equal to 100T.

Monticello Nuclear Generating Plant PTLR Revision 0 Page 7 of 38 The peak RPV ID fluence value of 6.43 x 101" 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].

2 The intenediate peak RPV ID fluence values of 2.77 x 1018 n/cm at 36 EFPY and 3.36 x 1018 n/cm 2 :at40 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 benctumarking 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 1/4T flaw. As a result, the 1/4T fluence for the limiting lower intermediate shell plate is 2.04 x 1018 n/cm 2 at 36 EFPY, 2.48 x 1018 n/cm 2 at 40 EFPY, and 4.75 x 1018 n/cm2 at 54 EFPY for MNGP.

The'RPV ID fluence value of 1.01 x 1018 n/cm 2 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 10' 7 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 1/4T flaw. As a result, the 1/4T fluence for the limiting recirculation inlet nozzle is 3.151 x 10"7 n/cm2 at 36 EFPY, 3.86 x 1017 n/cm2 at 40 EFPY, and 7.45 x 1.0' n/cin 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 lI4T location (inside surface flaw) and the 3/4T

Monticello Nuclear Generating Plant PTLR Revision 0 Page 8 of 38 location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inmer 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 copldown. 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 < 100°F/hr foi 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 cunot be maintained.

The initial RTNDT, the chemistry (weight-percent copper and nickel) and ART at the i/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10i7 h/cm2 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 0 Page 9 of 38 shell plate material in the vessel beltline region of MNGP. For plate heat C2220, since the scatter in the fitted results is less than 1-sigma (17TF), the margin term (aA = 17*F) is cut in half for the plate material when calculating the ART. Tile 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 RG1.99 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:

0 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/bh, which is

Monticello Nuclear Generating Plant PTLR Revision 0 Page 10 of 38 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 100*F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region.

o 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/h1r-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 synmmetric, 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 MNGP 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 folliwing 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 0 F/lhr shutdown transient [12]. Therefore, the stresses represent the bounding stresses in the recirculation inlet nozzle associated with i 00°Fihr heatup/cooldown limits associated with the P-T curves for a nozzle in the beltline region.

o Heat transfer coefficients were calculated in accordance with the MNGP recirculation inlet nozzle governing basis stress report. The heat transfer 1coefficients 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 0 Page 11 of 38 coefficient used the higher of the calculated vessel heat transfer coefficient (675 Btu/hr-f12 -F) or the calculated nozzle heat transfer coefficient (265 Btuh/u.'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 EPUvs. 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 Ouide 1.190 to determinie 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 hnd curves in the PTLR.

Monticello Nuclear Generating Plant PTLR Revision 0 Page 12 of 38 Excess Conservatism in Fluence and Multiple CUrvesfor HydrostaticPressure Test While reviewing the EPU-only flutence. calculation, it Was determined that fluence values for locations with an accumulated fluence nearer to the lower bound of 1.0 'x 1017 nl/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 212T.F. 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 0 Page 13 of 38 Component Fluencel[5 RPV Component 36 EFPY 40 EFPY 54 EFPY n/cm 2 n/cm2 n/cm2 Upper Intermediate Shell Plates 1.97x101" 2.30x101 4.06xl 0'7 (1-12 and 1-13)

Lower Intermediate Shell Plates 2 1 336x10'8 6.43x10 2.77x3 1 (1-14 and 1-15)

Lower Shell Plates (1-16 and 1-17) 1.85xl0M 8 2.28x10' 8 4.46x 10' 8 Liniting Weld 2.77x1 010 3.36x I 0'8 6.43x I 0" N-2 Nozzles 4.27x 10 7 5.23xl10' 1.0ixio08 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 300 Surveillance Capslde 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 0 Page 14 of 38 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 fl-tience calculation from 2007. [15]

Monticello Nuclear Generating Plant PTLR Revision 0 Page 15 of 38 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 File No. GE-10Q-401.

2. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods 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 Irtegrity Associates Calculation No. 1000847.303, Revision 2, "Revised P-T Curves Calculation," August 2011. (MNGP Site Calculation 11-005, Rev. OA)

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 0 Page 16 of 38

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

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Monticello Nuclear Generating Plant PTLR Revision 0 Page 22 of 38 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY Beltline Region P-T Curve P-T Curve Temperature Pressure 60.00 0 60.00 5.0 60.00 100 60.00 150 60.00 200 60.00 250 60.00 .300 60.00 312 60.00 313 61.75 350 86.10 400 102.40 450 114.67 500 1.24.52 550 132.74 600 139.81 650 146.00 700 151.49 750 156.45 800 160.96 850 165.09 900 168.91 950 172.47 1.000 175.78 1050 178.88 1100 8I1.82 1150 184.58 1200 187.19 1250 189.69 1300

Monticello Nuclear Generating Plant PTLR Revision 0 Page 23 of 38 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY (continued)

Plant = M*N*p Component = .Bottom Head (penetrations portion)

Bottom Head thickness, t = 5.938 inches Bottom'Head Radius, R = 1093.1875 inches ART = ~26.0 'F Kit = !

0.00 (no thermal effects)

Safety Factor = 1.50 Stress Concentration Factor = 3.00 (bottom head penetrations)

Mm= 2.256 Temperature Adjustment 0.0 'F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = 758.00 inches Pressure Adjustment = 27.4 psig (hydrostatic pressure head for a full vessel at 70*F)

Presure Adjustment = 0.0 psig (instrument uncertainty)

Gauge Adjusted Fluid Temperature Pressure for Temperature .KI, Kim for P-T Curve P-T Curve (ksi*inchl1 2 ) (ksi*inchll 2 ) (*F) (psig) 60.0 74.13 49.42 60 0 60.0 74.13 49.42 60 813 62.0 75.80 50.53 62 832 64.0 77.54 51.69 64 851 66.0 79.34 52.90 66 872 68.0 :81.23 54.15 68 893 70.0 83.19 55:46 70 915 72.0 85.23 56.82 72 939 74.10 87.35 58..23 74 963 76.0 89.56 59.71 76 988 78.0 91.86 61.24 78 1,014 80.0 94.25 62.84 80 1,041 82.0 96.75 64.50 82 1,069 84.0 99.34 66.23 84 1,099 86.0 102.04 68.03 86 1,129 88.0 104.85 69.90 88 1,161 90.0 107.77 71.85 90 1,194

.92.0 110.82 73.88 92 1,229 94.0 113.98 75.99 94 1,265 96.0 117.28 78.19 96 1,302

Monticello Nuclear Generating Plant PTLR Revision 0 Page 24 of 38 Table 1: MNGP Pressure Test (Curve A) P-T Curves for 36 EFPY (continued)

Plant =; -MNGP" -

Component = Uppar V6esel ART= 40.0 *F Vessel Radius R 103 i ncnes Nozzle comer. thickness t 7.732 1nches, approximate K11 =, 0.00k no thermnal effects) 69.10 lk sl'inch t/2 Crack Depth, a = 1.933 iriches Safety Fact~or= 1.60" Temperature Adjustment = 0.0 .'F (applied after bolt-up, Instrument uncertainty)

Height of Water for a Full Vessel = 788.00 Inches Pressure Adjustment = -27.4 ipsig (hydrostatic pressure head for a full essel at 70'F)

Pressure Adjustment = 0.0 psig instrument uncertainty)

Reference Pressure 1,000 fpsig (pressure at which the FEA stress coefficients are ealid)

Unit Pressure F 1,663 ipsig (hydrostatic pressure)

Flange RTNDT,= 'F ..... => All EFPY Gauge P-T P-T Curve Fluid Curve 10CFR5O Temperature K'p Tom porature Adjustments 6.F) (ksli'nch1 12 ) (ksl'lnch"*) ('F) (psig) 60.0 64.13 42.75 Tem 60 0 60.0 134.13 42.75 60 313 62.0 65,39 43.60 100 313 64.0 66.71 44.47 100 616 66.0 68.08 45.38 100 629 68.0 69.50 46.33 100 643 70.0 70.98 47.32 100 657 72.0 72.52 48.35 100 672 74.0 74.13 49.42 100 688 76.0 75.80 50.53 100 704 78.0 77.54 51.69 100 721 80.0 79,34 52.90 100 738 82.0 81.23 54.15 100 756 84.0 83.19 55.46 100 775 86.0 85.23 58.23 100 795 88.0 87.35 100 815 90.0 89.56 59,71 100 837 92,0 91.86 61.24 100 859 94.0 94.25 62.84 100 882 96.0 96.75 64.50 100 906 98.0 99.34 66.23 100 931 100.0 102.04 68.03 100 957 102.0 104.85 69.90 102 984 104.0 107.77 71.85 104 1012 106.0 110.82 73.88 106 1042 108.0 113.98 75.99 108 1072 110.0 117.28 78.19 110 1104 112.0 120.71 80.47 112 1137 114.0 124.28 82.86 114 1172 116.0 128.00 85.33 116 1207 118.0 131.87 87.91 118 1245 120.0 135.90 90.60 120 1284 122.0 140,09 93.39 122 1324 124.0 144.45 96.30 124 1366

Monticello Nuclear Generating Plant PTLR Revision 0 Page 25 of 38 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY Beltline Region P-T Curve P-T Curve Temperature Pressure 60.00 0 60.00 50 60.00 100 60.00 150 60.00 200 60.00 250 60.00 300 60.00 312 60.00 313 67.65 350 92.00 400 108.30 450 120.57 500 130.42 550 138.64 600 145;71 650 151.89 700 157.39 750 162.35 800 166.86 850 170.99 900 174.81 950 178.37 10oo 181.68 1050 184.78 1100 187.71 1150 190.48 1200 193.41 1250 196.73 1300

Monticello Nuclear Generating Plant PTLR Revision 0 Page 26 of 38 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY (continued)

Plant = MNGPj Component = Bottom Hea dc(penetrations portion)

Bottom Head thickness, t = ~5.938 inches Bottom Head Radius, R= 103.1 875 inches ART = oF Kit '0 00 i,(no thermal effects)

Safety Factor*=- 1.50 Stress Concentration-Factor = 3.00 (bottom head penetirations) mM = 2.256 Temperature Adjustment = 0.0 °F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = 758.00 inches Pressure Adjustment = 27.4 Spsig (hydrostatic pressure head for a full Vessel at 70'F)

Pressure Adjustment 0.0 psig (instrument uncertainty)

Gauge Adjusted Fluid Temperature Pressure for Temperature Kic Kim for P-T Curve P-T Curve (OF) (kslinchiZ') (ksl4inch..) (OF) (psIg) 60.0 74.13 49.42 60 " 0 60.0 74.13 49.42 60 813 62.0 75.80 50.53 62 832 64.0 77.54 51.69 64 851 66.0 79.34 52.90 66 872 68.0 81.23 54.15 68 893 70.0 83.19 55.46 915 85.23 ý70 72.0 56.82 72 939

74.0 87.35 58.23 ..74 963 76.0 89.56 59.71 76 988 78.0 91.86 61.24 78 1,014 80.0 94.25 62.84 80 1,041 82.0 96.75 64.50 :82 1,069 84.0 99.34 66.23 84 1,099 86.0 102.04 68.03 '86 1,129 88.0 104.85 69.90. 88 90.0 107.77 71.85 90 1,194 92.0 110.82 73.88 :92 1,229 94.0 .113.98 75.99 94 1,265 117.28 78.19 96 1,302 96.0

Monticello Nuclear Generating Plant PTLR Revision 0 Page 27 of 38 Table 2: MNGP Pressure Test (Curve A) P-T Curves for 40 EFPY (continued)

Plant= MNGP-Component = UpplerirVss ART=**'  :'F 40.0K Vessel Radius, R= 103 incnes Nozzle coiner thickness Vt' 7"732 inches, approximate 0.00  :(no thermal12 effects)

Kip.app~d 69.10, ksi*inch Crack Depth, a- 1.933 inches Safety Factor = 1.50.

Temperature Adjustment = 0.0 'F (applied after bolt-up, Instrument uncertainty)

Height of Water for a Full Vessel = ilnches Pressure Adjustment = 27.4 psig (hydrostatic pressure head for a full vessel at 70*F)

Pressure Adjustment = ~0.0'

.1,I000

psig (instrument uncertainty)

Reference Pressure *psig (pressure at which the FEA stress coefficients are valid)

Unit Pressure 1,68 3, :psIq (hydrostatic pressure)

Flange RTNODT=: 'F....> All EFPY Gauge P-T P-T Curve Fluid Curve 10CFRSO Temperature Kip Temperature Adjustments 12 F.0) (ks*inch' ) (kslinch"') (6) (psig) 60.0 64.13 42.75 60 0

60. 0 64.13 42.75 60 313 62.0 65.39 43.60 100 313 64.0 68.71 44.47 100 616 66.0 68.08 45.38 100 629 68.0 69.50 46,33 100 643 70.0 70.98 47.32 100 657 72.0 72.52 48.35 100 672 74.0 74.13 49.42 100 688 76.0 75.80 50.53 100 704 78.0 77.54 51.69 100 721 80.0 79.34 52.90 100 738 82.0 81.23 54.15 100 756 84.0 83.19 55.46 100 775 86.0 85.23 56.82 100 795 88.0 87.35 58.23 i00 815 90.0 89.56 59.71 100 837 92.0 91.86 61.24 100 859 94.0 94.25 62.84 100 882 96.0 96.75 64.50 100 906 98.0 99.34 66.23 100 931 100.0 102.04 ,68.03 100 957 102.0 104.85 69.90 102 984 104.0 107.77 71.85 104 1012 106.0 110.82 73.88 106 1042 108.0 113.98 75.99 108 1072 110.0 117.28 78.19 110 1104 112.0 120.71 80.47 112 1137 114.0 124.28 82.86 114 1172 116.0 128.00 85.33 116 1207 118.0 131.87 87.91 118 1245 120.0 135.90 90.60 120 1264 122.0 140.09 93.39 122 1324 124.0 144.45 96.30 124 1366

Monticello Nuclear Generating Plant PTLR Revision 0 Page 28 of 38 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY Beltline Region P-T Curve P-T Curve Temperature Pressure 60.00 0 60.00 50 60.00 100 60.00 150 60.00 200 60.00 250 60.00 300 60.00 312 60.00 313 82.85 350 107.19 400 W23.50 450 135.78 500 145.62 550 153.85 600 160.90 650 167.09 700 172.59 750 177.55 800 184.05 850 191.16 900 197.39 950 202.93 I6000 207.92 1050 212.45 1100 216.61 1150 220.44 1200 224.02 1250 227.33 1300

Monticello Nuclear Generating Plant PTLR Revision 0 Page 29 of38 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Plant =-'MNGP Component =Bottom Head (penetrations portion)

Bottom Head thickness, t = 5.938 inches Bottom Head Radius, R =

ART =

103.1875 26.0

'OF inches Kit (no thermal effects) 0.00 Safety Factor= 1.50 Stress Concentration Factor - ~3.00 (bottom head penetrations)

Mm = 2.256 Temperature Adjustment = ~0.0 *F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = V 758.00 Inches Pressure Adjustment [ ,psig (hydrostatic pressure head for a full vessel at 70*F)

Pressure Adjustment =  : psig (instrument uncertainty)

Gauge Adjusted Fluid Temperature Pressure for Kic Temperature KIM for P-T Curve P-T Curve (OF) ksi*i nch"' 2 ) (ksi*inchlI 2 ) (OF) (psig) 60.0 74.13 49.42 60 0 60..0 74.13 49.42 60 813 62.0 75.80 50.53 62 832 64.0 77.54 51.69 64 851 66.0 79.34 52.90 66 872 68.0 81.23 54.15 68 893 70.0 83.19 55.46 70 915 72.0 85.23 56.82 72 939 74.0 58.23 74 963 76.0 89.56 59.71 76 988

.78.0 91.86 61.24 78 1,014 80.0 94.25 62.84 80 1,041

.82.0 96.75 64.50 82 1,069 84.0 99.34 66.23 84 1,099 86.0 102.04 68.03 86 1,129 88.0 104.85 69.90 88 1,161 90.0 107.77 71.85 90 1,194 92.0 110.82 73.88 92 1,229 94.0 113.98 75.99 94 1,265 96.0 117.28 78.19 96 1,302

Monticello Nuclear Generating Plant PTLR Revision 0 Page 30 of 38 Table 3: MNGP Pressure Test (Curve A) P-T Curves for 54 EFPY (continued)

Plant =  :'.MIQNGP"':'4 Component= Upper Vse ART= .,40.0 , F Vessel Radius, R incnes Nozzle comer thickness, t'= -7.7'32 inches, approximate Kit '0060 (no thernal effects)

Kllpr,m,d *: 6.i0"tksi°inchII2 Crack SafetyDepth, a 69 I1klnchs 1 .933^ 1,inches Factor= . 180 Temperature Adjustment = 00 'F (applied after bolt-up, Instrument uncertainty)

Height of Water for a Full Vessel = 7 00 inches Pressure Adjustment 274 psig (hydrostatic pressure head for a fuillvessel at 70'F)

Pressure Adjustment = 0 .0 psig (instrument Uncertainty)

Reference Pressure -1,000, psig (pressure at which the FEA stress coefficients are valid)

Unit Pressure 1,563 psig (hydrostatic pressure)

Flange RTNDT 100* ==' > AIl:EFPY Gauge P-T P-T Curve Fluid Curve IOCFR60 Temrperature Temperature Adjustments 2

("F) (kSl*lnch 1 ) *(ksi*inch"') 0(F) (psig) 60.0 64.13 42.75 60 0 60.0 64.13 42.75 60 313 62.0 65.39 43.60 100 313 64.0 66.71 44.47 616 66.0 68.08 45.38 100 629 68.0 69.50 .46.3.3 100 643 70.0 .70.98 47.32 100 657 72.0 72.52 48.35 100 672 74.0 74.13 49.42 100 688 76.0 75.80 50.53 100 704 78.0 77.54 51.69 100 721 80.0 79.34 52.90 100 738 82.0 81.23 54.15 100 756 84.0 83.19 55.46 775 100 86.0 85.23 56.82 100 795 88.0 87.35 58.23 100 815 90.0 89,56 59.71 100 837 920 91.86 61.24 100 859 94.0 94.25 62.84 882 100 96.0 96.75 64.50 100 906 98.0 .99.34 66.23 100 100 931 100.0 102.04 68.03 957 102.0 104.85 69.90 102 984 104.0 107.77 71.85 104 1012 106.0 110.82 73.88 106 1042 108.0 113.98 75.99 .108 1072 110.0 117.28 78.19 110 1104 112.0 120.71 60.47 112 1137 114.0 124.28 82.86 114 1172 116.0 128.00 85.33 116 1207 118.0 131.87 87.91 118 1245 1120.0 135.90 90.60 120 128 122.0 140.09 93.39 122 1324 124.0 144.45 96.30 124 1366

Monticello Nuclear Generating Plant PTLR Revision 0 Page 31 of 38 Table 4: MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY Beltline Region P-T Curve P-T Curve Temperature Pressure 60.00 0 60.00 50 60.00 100 60.00 150 89.07 200 116.72 250 134.43 300 137.89 312 138.16 313 147.47 350 157.81 400 166,37 450 174.76 500 185.75 550 194.74 600 202.37 650 208.98 700 214.82 750 220.05 800 224.78 850 229.10 900 233.08 950 236.77 1000 240.21 1050 243.41 1100 246.43 1150 249.28 1200 251.98 1250 254.53 1300

Monticello Nuclear Generating Plant PTLR Revision 0 Page.32 of 38 Table 4: MNGP Core Not Critical (Curve B) P-T Curves for 54 EFPY (continued)

Plant = MMGP Component =Bottom Head (penetrations portion)

Bottom Head thickness, t= 5.938 inches Bottom Heaad Radius, R 103-1875 inches ART 26,0 T Kit 8 19 :.:ksinch Sfafety Factor.= 2.00 Stress Concentr. ation Factor = 3*00 (bottom head penetrat ions)

Mmi 2.256 Temperature Adjustment 0.0 'oF(applied after bolt-u p, instrumehntuncertainty)

Height of Water for a Full Vessel

. 758.00 .inches Pressure Adjustment 27.4 psig (hydrostatic pres=sure head for a full vessel at 70*F)

Pressure Adjustment - 0.0 psig (instrument unceirtainty)

Heat Up and Cool Down Rate - 100 *F/Hr Gauge Adjusted Fluid Temperature Pressure for Temperature KIC for P-T Curve P-T Curve (ksi*inch'1 2) (ksi*lnch1 2) (ipsig) 60.0 74.13 32.97 60 0 74,. 13 32.97 60 533 62.0 75.80 33.81 62 547 64.0 77.54 34.67 64 562 66.0 79.34 35.58 66 578 68.0 81.23 36.52 68 594 70.0 83.19 37.50 70 610 72.0 :85..23 72 628 74.0 87.35 39.58 .74 646 76.0 89.56 40.69 76 664 78.0 91.86 41.84 78 684 80.0 94.25 43.03 80 704 82.0 96.75 44.28 82 725 84 84.0 99.34 45.58 747 86.0 102.04 46.93 86 770 88.0 104.85 48.33 88 794 90.0 107.77 49.79 90 819 92.0 110.82 51.31 92 845

94. 0 113.98 52.90 94 872 96.0 117.28 54.55 96 900 98.0 120.71 56.26 98 929 100.0 124.28 58.05 100 960 99?

1o2.0 128.00 59.91 102 104.0 131.87 61.84, 104 1,024 106.0 135.90 63.85 106 1 058

.108.0 140.09 65.95; 108 1,094 110.0 144.45 68.13 110 1,131 112.0 148.99 70.40 11?

1,17.0 114.0 153.72 72.76 114 1,210 116.0 158.63 75.22 116 1,251 118&0 163.75 77.78 1.18 1,295 120.0 169.08 80.45" 120 1,340

Monticello Nuclear Generating Plant PTLR Revision 0 Page 33 of 38 Table 4: MNGP Core Not Critical (Curwe B) P-T Curves for 54 EFPY (continued)

Plant 1MNGP Component =UperýP6lVee ART.= ,400" 'F Vessel Radus. R = 103 tinches Nozzle comer thickness, t 7 .732 inches, approximate K ~

708 ý. ksi inch" Ki;., = 61 kslnihch'I Cra~ck Depth a= 103 Inches Sately Factor= 2,,00 Temperature Adjustment , 0 F (applied alter bolt-up. instrument uncertainty)

Height of Walerbfr a Full Vessel - 788 00 ,Iinches Pressure Adjustment , 2.74 ps!g (hydrostatic pressure head for a full wsset at 70TF) 1

'Pressure Adjustm ent = 0 i psig (Instrume6t uncertainty)

Reference Pressure ',1*,000 Ipsig (pressure at which the FEA etress coefficents are %slid)

Unit Pressure .s1,583 pg (hydrostatic pressure)

Flang RTIDr 160-0` -F = =' Att EFPY Gauge Fluid P-T P'T Curve Curve TemTIperature

('F) Y4. NO 0 Temperature Pressure (kst'inchrh) (ksi-lnch' ) (*F) (petg) 60.0 64.13 28.53 60 0 60.0 84.1,3 28:53 60 313 62.0 65.39 29.16 130 313 64.0 66.71 29.82 130 404 66.0 68.08 30.51 130 414 68.0 69.50 31.22 130 424 70.0 70.98 31.96 130 435 72.0 72.62 32.73 130 446 74.0 .74.13 33.53 130 458 76.0 75.80 34.37 130 470 78.0 77.54 35.24 130 483 80.0 79.34 36.14 130 496 82.0 81.23 37.08 130 509 84.0 63.19 38.06 130 523 86.0 85.23 39.08 130 538 88.0 87.35 40.14 130 554 90.0 89.56 41.25 130 570 92.0 91.86 42.40 130 586 94.0 94.25 43.60 130 603 96.0 96.75 44.84 130 622 98.0 99.34 46.14 130 640 100.0 I02.04 47.49 130 660 102.0 104.85 48.89 130 680 104.0 107.77 50.35 130 701 1080 110.82 51.88 130 7ý23 1068.0 113,98 53.46 130 746 110.0 117,28 55.11 130 770 112.0 120,71 56.82 130 795

.114.0 124,28 58.61 130 821

l16.0 128.00 60.47 130 848 118.0 131.87 62.40 130 876 120.0 135,90 64.42 130 905 122.0 140,09 66.51 130 935 124.0 144.45 68.69 130 967 126.0 148.99 70.96 130 1000 128.0 153.72 73.33 130 1034 130.0 158.63 75.78 130 1069 132.0 163.75 78.34 132 1106 134.0 169.08 81.01 134 1145 136,0 174.63 83.78 136 1185 138.0 180.40 86.67 138 1227 140.0 186.40 89.67 140 1270 142.0 192.66 92.80 142 1315

Monticello Nuclear Generating Plant PTLR Revision 0 Page 34 of 38 Table 5: MNGP Core Critical (Curve C) P-T Curves for 54 EFPY Plant={ MG Curve A Leak Test Temperature = 206.0, zOF Curve A Pressure = 1,025.0 <psig Unit Pressure =- 1,563 ipsig (hydrostatic pressure) 0 Flange RTNDT =¶ 10.0 F P-T Curve P-T Curve

'Temperature Pressure.

70.00 0 70.00 50 70.00 100 70.00 150 129.07 200 156.72 25.0 174.43 300 177.89 3.12 206.00 313 206.00 350 206.00 40.0 206.37 450 214.76 500 225.75 550 234.74 600 242.37 650 248.98 700 254.82 750 260.05 800 264.78 850 269.10 900 273.08 950 276.77 1000 280.21 1050 283.41 1100 286.43 1150 289.28 1200 291.98 1250 294.53 1300

Monticello Nuclear Generating Plant PTLR Revision 0 Page 35 of 38 Table 6: MNGP ART Calculations for 361EFPY

'Chemistry hiiti duetso1t?

~Description . Code No. . HeaitNo-.. fluxTYp-e;& Cot No. InitialIRTND'T2). ____

8 Fa.c~tor ART~ar, Marin-Terms ART 0F

-~Cu(ot% , (TF),

.Vw , "~ (F) 1,COF), (-F-)1 Upper/lnt Shell I-12

.1-. C2089-1 - 0.0 0.35 0.50 199.50 28.0 14.0 0.0 56.1 2613-1 CpIthil11 -27.0 0.35 0.49 198.25 :27.9 :13.9 0.0 82.7 LowerfInt Shell I5 LowerintShelll14 , - --. --C22201:

C22202 - - - - - - . 27.0 27.0' :_ . 0.16:: :0.64-016 0.64 180.007 180.00 '103A4 1034 8.5--

8.5 00 : :147

• 0.0 47.4 Lower Shell 1-16 A0946-1 - 27.0 0.14 0.56 98.20 47.3 17.0 0.0 108.3 Lower Shell 1-17 - C2193-1 - 0.0 0.17 0.50 118.50 57.1 17.0 0.0 91.1 4 -. C~~henis -: - AdjustesFoit

Descnption: C¢ode No. e NChemistry

~ eN. Hei' Feýrr1tdr

. 1Tux:Tyii&& Lot No. In itialI RTjr(F ______ Factor_ [ý .

-~~~~~ C(Wt%,) N(-t:%

N 0/. ~~?F

.-. (F F Limiting Weld - Beltline - I E8018N -65.6 0.10. .99 134.90 77.5 1 :20  ; 2.7. :7-:

- ~ '-~.'>--Chenisty Ad hustmnFor 114t Cheminstry ac.-

Descripton C No. Plate;Location l!iitirRTt;(F) 1,_CMtrms n .Factr A'DTj

-,BoundingN-2 Nozzle. E21VW Plate l-16/1-17 40.0 0.18 14190 .086' 32.1- 16.0.-. -0.0 =104.1

. - WallThickness (in) -, FluenceatlD Attenuation, 114t Fliwnce~atlj/4t. Fluence Factbr, FF 2

- ~ *-.Fu~ll _/4 t.Ž (hcii2 2Ki- jnlcni ) (.SOI!9 Upper/Int Shell 1-12 5.063 1.266 1.97E+17 0.738 1.454E+17 0.141 Upper/Int Shell 1-13 5.063 1.266 1.97E+117 *0.738 1.454E+17 0.141 Lowerlint Shell 1-14 5.063 1.266 2.77E+18 .0.738 2.044E+18 0.575 Lower/Int Shell 1-15 5.063 1.266 2.77E+18 0.738 2.044E+18 0.575 Lower Shell 1-164 5.063 1.266 1.85E+18 0.738 1.365E+18 .0.482 Lower Shell 1-17 5.063 1.266 1.85E+18 0.738 1.365E+18 0.482 Limiting Weld - Beltinie 5.063 1266 2.77E+18 .0.738 2.044E+18 0.575 n -e 5...... ...........................

.........063.1...... 4 *27E+1

.................... ....... 738 ................. ...1"1 E 1'+-17 3..... 226........

0".2,26

Monticello Nuclear Generating Plant PTLR Revision 0 Page '36of 38 Table 7: MNGP ART Calculations for 40 EFPY C..:". .. .. Chemis .AdjustinenftFor,1i4t,.

e o CodewNo. kelatlh& u eLot. Initiam"lRT (*F)' ART AFctor !MarginfTermf, M ARTi6T

>oY Cu /o N i Upper/lnt Shell 1-12 C2089-1 - 0.0 0.35 0.50 199.50 31.0 15.5 0.0 61.9 pper/I.........

n.. ................. ...... ... - ......................... ......... 270 0.35 . 49 9.8.:.2.5. . . 8 15.4 .0.0 88.6

• • -Shl 13 .. 6C2..6------- ...... ----- ...

Lower/lntShell 1-14-r 3 C2220-1 " - - 27;0 0.16 - 0.64 :K--180.00 :85*

8112.0- 0.0 -156.0:

Lowe*f/lItSfielll5MS C22202 . -- 27.0. 0.16 0.64 18000* *112.0-  :'8.5- 0.0 156.0.

Lower Shell 1-16 A0946-1 - 27.0 0.14 0.56 98.20 51.9 17.0 0.0 112.9 Lower Shell 1-7 C2193-1 - 0.0 0.17 0.50 118.50- 62.7 17.0 0.0 96.7

- -~ hemisry Q~~u~dieintssr1Iy Descriptionh -Code No. Heat No. Flux-Type & Lot No. lnitial RTNwr (*F) __ _ -,,

Factor, ARTimr !MariniTerM`sý ARTr

o*:  :.. -* ..

-

• i-/Z;- ,-  : .) C a% (wt ) ( O,.. a )*1b (.F).

.F).

Lmiting0Weld E8018N -65.6 0.10 0.999Beltine134.90 83.9- 28.0 12.7 79.8 Chmsl Ctierfiiity 'Adjiju~.e ntsl~pbr.1Iujt_

Description CodeNo. -H~ht No-.- 0Plte"Loatibn~ 'niiiai TUm F _______-Fco &~~j agnTrsATu

- ,C w% Ni (wt%.- -OF ~0F! cr d1~ý~ORF) 10CFY, Bounding N-2 Nozzle - E2IVW Plate'l-161-17 -40.0& 0.18 -0.86 .141.90 36.0 :17:0 -0.0U 110.0:

-- -~ ~- ~. ,,

Wall'Thickness(in),:V

.~ Fluence ,Data,--ý,

FlueceatlD' Attenuation, 1/4t .Fluence~at-1/4t, FluenceFactOrFF

- - , FLocatior`D

- -- -'O2& -.-. -

ru11 , 114t- jrim)J.-, i--(h/cm iY Upper/Int Shell 1-12 5.063 1.266 2.30E+17 0.738 1.698E+17 0.155 Upper/Int Shell 1-13 5.063 1.266 2.30E+17 0.738 1.698E+17 0.155 Lowerlnt Shell 1-14 5U063 :1.266 3.36E+18 0,738 2.48E'+18& 0.622 Lower/Irt Shell 1-15 5.063 1.266 3.361E+18 0.738 2.48E+18 0.622 Lower Shell 1-16 5.063. 1.266 2.28E+18 0.738 1.683E418 0.529 Lower Shell 1-17 5.063 1.266 2.28E--18 0.738 1.683E4-18 0.529 Limiting Weld - Beltline ' 5.063 1.266 ý3.36E+18 0.738 2.48E+18 0.622 Bounding N-2 Nozzle 5.063 1.266 5.23E+17 0,738 3.86E+17 0.254

Monticello NuClear GeneratingPlant PTLR Revision 0 Page 37 of 38 Table 8: MNGP ART Calculations for 54:EFPY Descriptin jChemistry Cod No. eat.o..Flux Type .. & Lot'NoesInitial helY M T

______ _______Cu,(wt%) ~N Kr a,(_

Upper/mnt Shell 1-12 C2089-1 0.0 0.35 0.50 199.50 4318 17.0 0.0 77.8

.Upper.ntShel.-3.. -. C2613-1 27.0 0.35 0.49 198.25 .43.5 17.0 0.0 104.5 l~ower Shelll-14. -,.... C2220-2-* -27.0. .016..180.007 - 0.64 -: 8  :-142:6 0.0 186.6 Low'erllnt.Sheiel ......... ... 27.0, 016 - .4 100 142.,6 -8.5ý 1-00'-1865.

Lower Shell 1-16 A0946-1 27.0 0.14 0.56 98.20 68.2 17.0 0.0 129.2 Lower Shell 1-17 C21931. 0.0 0.17 0.50 118.50 82.3 .17.0 0.0 116.3ý

- i,4 Ch Ajusei~i~t§'F~or1~t:

Description- Co-deNo.0 Heat-No.. lux Type &.L*otfNo,'

. nitial RT'ryr..(OF)J _-__" _____ " Factor .. RT,6j IMarginTeh1,ARTN Flu~~~~ .- ,,Tco ARNi

.ý~oMrgnThr, 2 7 1202.8 D

Limiting Weld - Beltline - - E8018N, - -65.6 0.10 099: :134.90:1'1 106,'9 Chemistryt thryit~~dutrinsd~

tztz Description. o.e No.. Heat No. Plate Locationf InitialRTND ("17F) Factor.

-Ariit,kT,6rRIThm argin Terns

_________~~C __],__aA (________ (OF) a,__

Bounding N-2 Nozzle E21VW Plate 1-1611-17 40.0- 0.18 0.86 . 141.90 512.2 17.0 0.0 125.2

- Fluence Data,k  :, ' '

," WallThickness (in) Fluenceat ID, Attenuation, l1/4tta ';'ý.1Fluenceat 1/4t-. FluenceFactor, Funea FF.-ý Lo'cationh alTikns i)

UpperllIntShell 1-12 5.063 1.266 4.06E+17 0.738 .2.996E+17 0,219 Upper/IAnt Shell 1-13 5.063 :1.266 4.06E+17 0.738 2.996E+17 0.219 Lowe&int Shell 1-14 5.063 1.266 .643E+18 0.738 4.746E+18 0:792 Lower/nt Shell 1-15 5.063: 1266 .6.43E+18 0.738 4.746E+18. 0.792 Lower Shell1-16 :5.063 1.266 4A46E+18 0.738 3;292E+18 0.694 Lower Shell 1-17 5:063 1.266' 4.46E+18 0.738 3.292E+18 0.694 Limiting Weld - Beltline 5.063 1.266 .6.43E+18 0.738 4.746E+18 0.792

Bounding N-2 Nozzle 5.063 1.266 1.01 E+18 0.738 7.454E+17 0.361

Monticello Nuclear Generating Plant PTLR Revision 0 Page 38 of 38 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 tuiaxial 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 Progprams, and has been approved by NRC. Xcel Energy com*itted 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 ttest specimens fabricated using materials from the vessel materials within the core beitline 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.