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 S 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: Technical Review: Program Manager: Wynter McGruderýirn Bi mn-an Date: Date: 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/cm 2 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 intenediate peak RPV ID fluence values of 2.77 x 1018 n/cm 2 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/cm 2 at 36 EFPY, 3.86 x 1017 n/cm 2 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 > 10i 7 h/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 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-ft 2-°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.'ft 2 -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 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 Hydrostatic Pressure 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/cm 2 n/cm 2 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 (1-14 and 1-15) 2.77x3 1 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 5.23xl 10' 1.0ixio0 8 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 1 1-003, 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 U-w w w LU w-n w U, U)300 280 260 240 220 200 180 160 140 120 1100 80 60 40.20 0 COMPLIANCE REQUIRES OPERATION ABOVETHE CU RVES...............I I.I I I I .1 I.1 .I I .1 I I.I I.rA c-I M II 111 11 M i ll 111111 i u 4 i i ii i 11 1 1 1 1 Ill Ill~ III!i i..I. I I Bolt-up Temp: -Beitline Region 6OF --Bottorm.Head

,.Upper Vessel 0 CD 0 CD CD CD 0 100 200 300 400 500 600 700 800 900 PRESSURE LIMIT IN REACTOR VESSEL (psig)1000 1100 1200 1300 LJ-w w fn-j ul 0 I-300 280 260 240 220 200 180 160 140 120 100 80 60 40 M LACERI I RE OEAIO ABV TH URE IL~ ~ ~ iI~i I II T1 --- T~0 C,, U, U,-o 0 L L II1 11I 1.1 ,l1 _ : 1. ...1 1 1: 1 1 H i l 1 1- ' " I 1 1 1 "'-Beltline Region Bolt 60'F -8ottom Head........ ......... Upper-Vessel F illI ! ! i! [ i : !l l~20-0 CD CD 2)-v 0 100 200 300 400 500 600 700 800 900 PRESSURE LIMIT IN REACTOR VESSEL (psig)1000 11i00 1200 1300 00 00 300]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 1 280 j-Ianl wTh~I I I I I I I I i 1 ' ýCOMPLIANCE REQUIRES OPERATION ABOVETHECURVES 1~11~ II III.f. .. .... Il .......-1 .w 240~220 Lu 200_j, 180'u160 S140> 120 0 100 S 60 S40 20 0 I-1 1 1 1 1 .1 1 1 1 1 1 1 I l l ..... ..........-=El H.1:1= 1 [1 ciA LiJ IA 0A I II III I I I liii II II 111111 I I liii III I.1 4I 1111 1 1- 1 1 1 ' MT Bolt-up Tem~p: I -eltlineReg ion 60 !F -- Bottom Head S..... UpperVessel 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 0 CD)tn. -PRESSURE LIMIT IN REACTOR VESSEL (psig)0J 00 300 280 260 240 220 (L 200 LUJ 180 Lu 160 LU cn 140 cf, Lu 120 0~3100 Lu w 80" 60 40 20 0 C MPLIANCE REQUIRES OPERATION ABOVETHECURVES

-1 H 111. -W 111111114-4-4 70: 0-PEU LII I R I T.....Bottom Head.Bolt-up Temp:= * ...*Upper Vessel.. .0 100 200 300 .400 Soo 600 700 800 900 1000 1100 1200 1300 PRESSURE LIMIT INREACTOR VESSEL TOP HEAD (psig)0 0o (C)0-o..0 0 0 00 u-0.LUI LI-LI-LU LUI 0 F-C.300 280 260 240 220 200 180 160 140 120 100-I I I I k i 1.1. i] 11111.111 I 111111111 II liii. 'I II I~.I liii .1 .I..I.-I 111111 111.1.11111 I I T T I I I ..... ....... .COMPLIANCE REQUIRES OPERATION ABOVETHE CURVES L mi imu-----------riticalityTemp:

70*F------ ----J H LI-L I I I Ll Jmm It 0o 80 60 40 20 0)CD CDa 0 0 100 200 300 400 500 600 700 800 900 1000 PRESSURE LIMIT IN REACTOR VESSEL-TOP HEAD (psig)1100 1200 1300.0 00 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 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 1.24.52 132.74 139.81 146.00 151.49 156.45 160.96 165.09 168.91 172.47 175.78 178.88 8I1.82 184.58 187.19 189.69 P-T Curve Pressure 0 5.0 100 150 200 250.300 312 313 350 400 450 500 550 600 650 700 750 800 850 900 950 1.000 1050 1100 1150 1200 1250 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 = Component

= .Bottom Head Bottom Head thickness, t =Bottom'Head Radius, R =ART =Kit = !Safety Factor =Stress Concentration Factor =5.938 1093.1875~26.0 0.00 1.50 3.00 (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)

Mm= 2.256 Temperature Adjustment Height of Water for a Full Vessel =Pressure Adjustment

=Presure Adjustment

=0.0 758.00 27.4 0.0 Gauge Fluid Temperature Temperature 60.0 60.0 62.0 64.0 66.0 68.0 70.0 72.0 74.10 76.0 78.0 80.0 82.0 84.0 86.0 88.0 90.0.92.0 94.0 96.0.KI, (ksi*inchl1 2)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 Kim (ksi*inchll 2)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 for P-T Curve (*F)60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 Adjusted Pressure for P-T Curve (psig)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 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 Vessel Radius R 103 i Nozzle comer. thickness t 7.732 1 K 1 1=, 0.00k*F ncnes nches, approximate no thermnal effects)sl'inch t/2 riches Crack Depth, a =Safety Fact~or=Temperature Adjustment

=Height of Water for a Full Vessel =Pressure Adjustment

=Pressure Adjustment

=Reference Pressure Unit Pressure F Flange RTNDT,=Gauge Fluid Temperature 6.F)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 69.10 k l 1.933 1.60" 0.0 788.00-27.4 0.0 1,000 1,663 (ksli'nch 1 1 2)64.13 134.13 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.'F (applied after bolt-up, Instrument uncertainty)

Inches ipsig (hydrostatic pressure head for a full essel at 70'F)psig instrument uncertainty) fpsig (pressure at which the FEA stress coefficients are ealid)ipsig (hydrostatic pressure)'F .....=> All EFPY K'p 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 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 Tem i Tom P-T P-T Curve Curve 1OCFR5O porature Adjustments

('F) (psig)60 0 60 313 100 313 100 616 100 629 100 643 100 657 100 672 100 688 100 704 100 721 100 738 100 756 100 775 100 795 100 815 100 837 100 859 100 882 100 906 100 931 100 957 102 984 104 1012 106 1042 108 1072 110 1104 112 1137 114 1172 116 1207 118 1245 120 1284 122 1324 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 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 10oo 1050 1100 1150 1200 1250 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 Bottom Head thickness, t =Bottom Head Radius, R=ART =Kit Safety Factor*=-Stress Concentration-Factor

=~5.938 103.1 875'0 00 1.50 3.00 dc (penetrations portion)inches inches oF i, (no thermal effects)(bottom head penetirations)

°F (applied after bolt-up, instrument uncertainty) inches Spsig (hydrostatic pressure head for a full Vessel at 70'F)psig (instrument uncertainty) mM =Temperature Adjustment

=Height of Water for a Full Vessel =Pressure Adjustment

=Pressure Adjustment 2.256 0.0 758.00 27.4 0.0 Gauge Fluid Temperature (OF)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 Kic (kslinchiZ')

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 Kim (ksl4inch..)

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 Temperature for P-T Curve (OF)60 " 60 62 64 66 68ý70 72..74 76 78 80:82 84'86 88 90:92 94 96 Adjusted Pressure for P-T Curve (psIg)0 813 832 851 872 893 915 939 963 988 1,014 1,041 1,069 1,099 1,129 1,194 1,229 1,265 1,302 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

= UpplerirVss40.0K Vessel Radius, R= 103 Nozzle coiner thickness Vt' 7"732 0.00 Kip.app~d Crack Depth, a-Safety Factor =Temperature Adjustment

=Height of Water for a Full Vessel =Pressure Adjustment

=Pressure Adjustment

=Reference Pressure Unit Pressure Flange RTNODT=: Gauge Fluid Temperature F.0)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 69.10, 1.933 1.50.0.0 27.4~0.0.1,I000 '1,68 3, (ks*inch'1 2)64.13 64.13 65.39 68.71 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:'F incnes inches, approximate

(no thermal effects)ksi*inch 1 2 inches'F (applied after bolt-up, Instrument uncertainty) ilnches psig (hydrostatic pressure head for a full vessel at 70*F):psig (instrument uncertainty)(pressure at which the FEA stress coefficients are valid):psIq (hydrostatic pressure)'F....> All EFPY Kip (kslinch"')

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 P-T Curve Temperature (6)60 60 100 100 100 100 100 100 100 100 100 100 100 100 100 i00 100 100 100 100 100 100 102 104 106 108 110 112 114 116 118 120 122 124 P-T Curve 1OCFRSO Adjustments (psig)0 313 313 616 629 643 657 672 688 704 721 738 756 775 795 815 837 859 882 906 931 957 984 1012 1042 1072 1104 1137 1172 1207 1245 1264 1324 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 Temperature 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 60.00 82.85 107.19 W23.50 135.78 145.62 153.85 160.90 167.09 172.59 177.55 184.05 191.16 197.39 202.93 207.92 212.45 216.61 220.44 224.02 227.33 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 I6000 1050 1100 1150 1200 1250 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 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 = V Pressure Adjustment

[Pressure Adjustment

=Gauge Fluid Temperature (OF)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 5.938 103.1875 26.0 (penetrations portion)inches inches'OF 0.00 1.50~3.00 2.256~0.0 758.00 Kic ksi*i nch"'2)74.13 74.13 75.80 77.54 79.34 81.23 83.19 85.23 89.56 91.86 94.25 96.75 99.34 102.04 104.85 107.77 110.82 113.98 117.28 (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)

KIM (ksi*inchlI 2)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 Temperature for P-T Curve (OF)60 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 Adjusted Pressure for P-T Curve (psig)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 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 69 I 1klnchs Crack Depth, a 1 .933^ 1,inches Safety 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 fuill vessel 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 ==' > AIl:EFPY Gauge Fluid Temrperature

("F)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 90.0 920 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 1120.0 122.0 124.0 (kSl*lnch 1 2)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*(ksi*inch"')

42.75 42.75 43.60 44.47 45.38.46.3.3 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 60.47 82.86 85.33 87.91 90.60 93.39 96.30 P-T P-T Curve Curve IOCFR60 Temperature 0(F)60 60 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 102 104 106.108 110 112 114 116 118 120 122 124 Adjustments (psig)0 313 313 616 629 643 657 672 688 704 721 738 756 775 795 815 837 859 882 906 931 957 984 1012 1042 1072 1104 1137 1172 1207 1245 128 1324 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)

Bottom Head Bottom Hea Sf Stress Concentr.Temperature Height of Water for a Pressure Pressure Heat Up and Cool Plant = MMGP Component

=Bottom Head (penetrations portion)thickness, t= 5.938 inches ad Radius, R 103-1875 inches ART 26,0 T Kit 8 19 :.:ksinch afety Factor.= 2.00 ation Factor = (bottom head penetrat Mmi 2.256 Adjustment 0.0 'oF (applied after bolt-u Full Vessel .758.00 .inches Adjustment 27.4 psig (hydrostatic pres=Adjustment

-0.0 psig (instrument uncei Down Rate -100 *F/Hr ions)p, instrumehntuncertainty) sure head for a full vessel at 70*F)rtainty)Gauge Fluid Temperature 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 1o2.0 104.0 106.0.108 .0 110.0 112.0 114.0 116.0 118&0 120.0 KIC (ksi*inch' 1 2)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 (ksi*lnch1 2)32.97 32.97 33.81 34.67 35.58 36.52 37.50 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" Temperature for P-T Curve 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 11?114 116 1.18 120 Adjusted Pressure for P-T Curve (ipsig)0 533 547 562 578 594 610 628 646 664 684 704 725 747 770 794 819 845 872 900 929 960 99?1,024 1 058 1,094 1,131 1,17.0 1,210 1,251 1,295 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;., = 6 1 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 2.7 , 4 ps!g (hydrostatic pressure head for a full wsset at 70TF)'Pressure Adjustm ent = 0 i 1 psig (Instrume6t uncertainty)

Reference Pressure Ipsig (pressure at which the FEA etress coefficents are %slid)Unit Pressure .s1,583 pg (hydrostatic pressure)Flang RTI Dr 160- 0` -F = =' Att EFPY Gauge P-T P'T Fluid Curve Curve TemTIperature Y4. NO Temperature Pressure ('F) (kst'inchrh) (ksi-lnch'

0) (*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)Flange RTNDT =¶ 10.0 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. In itialIRTND'T2).

____ Fa.c~to r ART~ar, Marin-Terms ART-~Cu(ot% .Vw 8 , (TF), , 0 F "~ (F) 1,C OF), (-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 CpIthil11 2613-1 -27.0 0.35 0.49 198.25 :27.9 :13.9 0.0 82.7 LowerintShelll14 , --. --C22201: --.27.0 .016 :0.64- 180.007 1034 8.5-- :147 LowerfInt Shell I5 -C22202 ----27.0' :_ 0.16:: 0.64 180.00 '103A4 8.5 00 : 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 Feýrr1tdr

~ eN. Hei' .1Tux:Tyii&&

Lot No. In itialI RTjr(F Factor_ ______ [ý .-~~~~~ C(Wt%,) N N(-t:% 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,_ n .Factr CMtrms A'DTj-,Bounding N-2 Nozzle. E21VW Plate l-16/1-17 40.0 0.18 .086 14190 ' 32.1- 16.0.-. -0.0 =104.1.-WallThickness (in) -, FluenceatlD Attenuation, 114t Fliwnce~atlj/4t.

Fluence Factbr, FF-~ *-.Fu~ll _/4 t. (hcii 2 2Ki- jnlcni 2) (.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 1-1 7 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 .................

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

0".2,26 .....226........

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

e o CodewNo. kelatlh& u eLot. Initiam"lRT

(*F)' AFctor ART M !MarginfTermf, ARTi6T Cu /o N i >oY 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 8112.0- 0.0 -156.0:

C22202 .-- 27.0. 0.16 0.64 :'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 Description h -Code No. Heat No. Flux- Type & Lot No. lnitial RTNwr (*F) -,, __ _ Factor, ARTimr !MariniTerM`sý ARTr: : .. .. ... --i-/Z;- ,- : (wt C a% ) ( O, .) a b (.F). .. .F).Lmiting0Weld 9Beltine E8018N -65.6 0.10 0.99 134.90 83.9- 28.0 12.7 79.8 Chmsl Ctierfiiity

'Adjiju~.e ntsl~pbr.1Iujt_

Description CodeNo. -H~ht No-.- 0 Plte"Loatibn~

'niiiai TUm F _______-Fco

&~~j agnTrsATu-,C w% Ni (wt%.- ~ -OF cr 0 F! d1~ ý~ ORF) 10CF Y, Bounding N-2 Nozzle -E2IVW Plate'l-161-17

-40.0& 0.18 -0.86 .141.90 36.0 U :17:0 -0.0 110.0:-- -~ ~- ~. ,, .~ Fluence ,Data,--ý, Wall'Thickness(in),:V 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 jChemistry helY Descriptin Co d No. ..eat.o..Flux Type & Lot'Noes Initial 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.

-,....

-27.0. .016. -0.64 -: .180.007 :-142:6 8 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

.nitial RT'ryr (OF)J .. _-__" _____ " Factor ..RT,6j IMarginTeh1,ARTN Flu~~~~ .- ,,Tco .ý~oMrgnThr, ARNi Limiting Weld -Beltline --E8018N, --65.6 0.10 099: :134.90:1'1 106,'9 12 D 2 7 02.8-" '. , Chemistryt thryit~~dutrinsd~

Description.

o.e No.. Heat No. Plate Locationf InitialRTND

("17F) Factor. kT,6rRIThm argin Terns -Ariit,_________~~C

(________

__],__aA (OF) a,__ tztz 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, l1/4t ';'ý.1Fluenceat 1/4t-. FluenceFactor, FF.-ýLo'cationh alTikns i) ta Funea 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 Shell 1-16 :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 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.