ML100700088

Pressure and Temperature Limits Report Revision

ML100700088
Person / Time
Site:	Nine Mile Point
Issue date:	03/03/2010
From:	Joseph Pacher Constellation Energy Group, Nine Mile Point
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100700088 (26)

Text

IA:.

p.

CENG a joint venture of

&0t Constellation P.O. Box 63 Lycoming, NY 13093 NINE MILE POINT NUCLEAR STATION March 3, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:

SUBJECT:

Document Control Desk Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Pressure and Temperature Limits Report Revision Enclosed is a copy of the Pressure and Temperature Limits Report, PTLR-1, Revision 01.00, for Nine Mile Point Unit 1 (NMP1). This report revision is being submitted pursuant to NMP1 Technical Specification 6.6.7.c.

Should you have any questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.

ry truly yours, Joseph E. Pacher Manager Engineering Services JEP/DEV

Enclosure:

Nine Mile Point Unit 1 Pressure and Temperature Limits Report, PTLR-1, Revision 01.00 cc:

S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC Aoo(

1J0 (L_-

ENCLOSURE NINE MILE POINT UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT PTLR-1, REVISION 01.00 Nine Mile Point Nuclear Station, LLC March 3, 2010

NMP1 Pressure and Temperature Limits Report 0

Constellation Energy, Nine Mile Point Nuclear Station NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 1 Pressure and Temperature Limits Report (PTLR)

PTLR-1, Revision 01.00 Prepared by:

Reviewed by:

Approved by:

Approved by:

R. Corieri Mech./Struc. Design Engineering G. B. Inch Mech./Struc. Design Engineering P. E/tartolini, Supervisor Mech./Struc. Design Engineering A. D. Sterio, General Supervisor Design Engineering Date:

,ý& //'o Date:

2i1V10 Date:

Date: 2 IL *94 This Controlled Document provides reactor pressure vessel pressure and temperature limits for use in conjunction with the Nine Mile Point Unit I Technical Specifications. Document pages may only be changed through the re-issue of a revision to the entire document.

NMPI Pressure and Temperature Limits Report Section 1.0 2.0 3.0 4.0 5.0 6.0 Table of Contents Title Purpose Applicability Methodology Operating Limits Discussion References NMP1 Pressure Test (Curve A)

NMP1 Normal Operation (Heatup and Cooldown) - Core Not Critical (Curve B)

NMP1 Normal Operation (Heatup and Cooldown - Core Critical (Curve C)

NMP1 Feedwater Nozzle Finite Element Model NMP1 Pressure Test (Curve A) - Beltline Region, EOC 22 NMP1 Normal Operation - Core Not Critical (Curve B), Beltline Region, EOC 22 NMP1 Normal Operation - Core Critical (Curve C), EOC 22 NMP1 ART Calculations for 36 EFPY Heat Transfer Coefficients for NMP1 Feedwater Nozzle Feedwater Nozzle Material Properties Page 1

1 1

2 3

7 Figure 1 Figure 2 Figure 3 Figure 4 Table 1 Table 2 Table 3 Table 4 Table 5 Table 6 9

10 11 12 13 15 17 19 20 21 22 Appendix A NMP1 Reactor Vessel Materials Surveillance Program PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report 1.0 PURPOSE The purpose of the Nine Mile Point Nuclear Station Unit 1 (NMP1) Pressure and Temperature Limits Report (PTLR) is to present operating Iimits relating to:

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

2. RCS Heatup and Cooldown rates.

3. Reactor Pressure Vessel (RPV) head flange bolt-up temperature limits.

This report has been prepared in accordance with the requirements of Technical Specification (TS) Section 6.6.7, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," and the template provided in Licensing Topical Report SIR-05-044, Revision 0 (Reference 6.1).

2.0 APPLICABILITY This report is applicable to the NMP1 RPV until the end of operating cycle 22. The following TS sections are affected by the information contained in this report:

Limiting Condition for Operation Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates."

Limiting Condition for Operation Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

Surveillance Requirement Section 4.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

3.0 METHODOLOGY The limits in this report were derived as follows:

(1) The methodology used to calculate the pressure and temperature limits is in accordance with Reference 6.1, which has been approved for BWR use by the NRC. The pressure and temperature limit calculations are documented in Reference 6.2.

Page 1 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report (2) The neutron fluence is calculated in accordance with NRC Regulatory Guide (RG) 1.190 (Reference 6.3) based on the wetted surface fluence that is documented in Reference 6.20. The methodology used to calculate the RPV neutron fluence has been approved by the NRC in Reference 6.5.

(3) The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (Reference 6.6),

as documented in Reference 6.7 as amended by Reference 6.21.

(4) This revision of the pressure and temperature limits is to incorporate the following changes:

Rev. 0 - Initial issue of PTLR.

Rev. 01.00:

o Removed 28 EFPY curves/tables which are no longer valid o

Limited the use of the 36 EFPY curves/tables until the end of operating cycle (EOC) 22 o

Removed 46 EFPY curves/tables which are not used during the current fluence period o Added a footnote 4 and clarification to fourth bullet in Section 4.0 to make the described operating limit consistent with the applicable PTLR Figure.

o Previous Technical Specification Pressure-Temperature Limit Figures only had one curve per Figure. The PTLR Figures include several curves. A note was added to the bottom of each PTLR figure indicating plant operation shall remain to the right of all of the curves shown in each Figure.

o Changed the legend for the temperature axis on each curve from "metal" temperature to "coolant" temperature to be consistent with the sample Figures 2-2 and 2-3 in Reference 6.1.

o Added a footnote 1 to the last two columns of the Curve A and B Tables to make more evident that the data in the last two columns was used to plot the respective curves with instrument uncertainty and static head corrections added.

o Removed the 28 EFPY and 46 EFPY ART Calculation Tables corresponding to the removal of the 28 EFPY and 46 EFPY PT curves/tables.

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, in accordance with TS Section 6.6.7.

4.0 OPERATING LIMITS The pressure-temperature (P-T) limit curves included in this report represent steam dome pressure versus minimum vessel coolant temperature (as measured from recirculation loop Page 2 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report suction) 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 (heatup and cooldown), referred to as Curve B; and (c) core critical operation (heatup and cooldown), referred to as Curve C.

Complete P-T limit curves were developed for 28, 36 and 46 EFPY for NMP1, as documented in Reference 6.2. Only the NMP1 P-T limit curves for the current fluence period are included in this report. The applicable NMP1 P-T limit curves for this fluence period are included in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3.

Other conditions applicable to the NMP1 RPV are:

Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figure 1:

Curve A): 5 25°F/hourl.

Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): :5 100°F/hour2.

RPV head installation temperature (i.e., bolt-up) and core not critical limit (Figure 1:

Curve A - Hydrostatic and Class 1 Leak Testing; Figure 2: Curve B - non-nuclear heating): > 70=F.

RPV flange and adjacent shell temperature core critical limit (Figure 3: Curve C -

nuclear heating): _> 100°F 4.

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 (Reference 6.6) provides the methods for determining the ART. The RG 1.99 methods for 1 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 1-hour period is less than or equal to 100°F.

3 A higher minimum bolt-up temperature of 70'F was applied to these curves, as compared to the 60OF value determined in Reference 6.2, in order to be consistent with the minimum bolt-up temperature value used in previous studies.

4With water level within the normal range for power operation, the minimum criticality temperature of 100lF is determined from the RTNDT of the closure flange region + 60'F.

Page 3 of 22 PTLR-1

.Revision 01.00

NMP1 Pressure and Temperature Limits Report 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 NMP1 vessel plate and weld materials (Reference 6.7). The Cu and Ni values were used with Tables 1 and 2 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds and plates, respectively.

The peak RPV inside diameter (ID) fluence values of 1.12 x 1018 n/cm2 at 28 EFPY and 1.61 x 1018 n/cm2 at 46 EFPY used in the P-T curve evaluation were obtained from Reference 6.4.

Neutron fluence values were calculated using methods that conform to the guidelines of RG 1.190 (Reference 6.3). At 36 EFPY, the peak fluence value of 1.34 x 1018 n/cm2 was obtained by performing a linear interpolation between the fluence values at 28 and 46 EFPY. These fluence values apply to the limiting beltline lower shell plate (Heat No. P2112 for NMP1). The fluence values for the lower shell plates are based upon an attenuation factor of 0.652 for a postulated 1/4T flaw. As a result, the 1/4T fluences for 28, 36 and 46 EFPY for the limiting lower shell plate are 7.29 x 1017, 8.71 x 1017 and 1.05 x 1018 n/cm2, respectively, for NMP1. The peak RPV ID fluence values described above were revised in Reference 6.20 after Revision 0 of this PTLR was submitted to the NRC and prior to Implementation of Technical Specification, Amendment # 204. Because the revised peak RPV ID fluence values increased, the validity period for the 36 EFPY curves has been shortened until the end of operating cycle 22 (< 36 EFPY) as documented in Reference 6.22. At the end of cycle 22, the peak fluence used to develop the 36. EFPY curve is conservatively predicted to be equivalent to the revised 36 EFPY peak fluence documented in Reference 6.20.

The P-T limit curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stresses at the 114T location are assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T location to be less than that at the Page 4 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report 3/4T location for a given metal temperature. This approach causes no operational difficulties, since the boiling water reactor 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 are applicable for a coolant heatup and cooldown temperature rate of _ 1O0°F/hr. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (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. Thus, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial nil-ductility transition reference temperature (RTNDT), the chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 101 7 n/cm2 for E > 1MeV) are shown in Table 4 for 36. EFPY, based on Reference 6.7. The initial RTNDTvalues were determined and reported to the NRC in the NMP1 responses to NRC Generic Letter (GL) 92-01, Revision 1 (Reference 6.8) and GL 92-01, Revision 1, Supplement 1 (Reference 6.9). The NRC acknowledged these GL responses in letters dated March 30, 1994, August 26, 1996, and June 25, 1999 (References 6.10, 6.11, and 6.12, respectively). The initial RTNDT values shown in Table 4 have previously been used in establishing the current TS P-T limit curves (license amendment approved by the NRC in Reference 6.5) and in evaluations contained in the License Renewal Application (approved by the NRC in Reference 6.13).

Per Reference 6.7 and in accordance with Appendix A of Reference 6.1, the NMP1 representative weld and plate surveillance materials data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) were reviewed. The representative heats of plate materials (P2112 and P2130) in the ISP are the same as the lower shell plate material in the vessel beltline region of NMP1. For plate heat P2112, since the scatter in the fitted results exceeds 1-sigma (17iF), the full 2-sigma margin term has been utilized in calculating the ART value for this plate in the vessel. For plate heat P2130, since the surveillance data was found to be credible, the margin term (oA = 17'F) is divided by two for the Page 5 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report plate material when calculating the ART. Therefore, the CFs from the NRC's Reactor Vessel Integrity Database (Reference 6.14) and Reference 6.6 were used in the determination of ART for all NMP1 materials except for plate heat P2130.

The only computer code used in the determination of the NMP1 P-T curves was the ANSYS/Mechanical Release 6.1 (with Service Packs 2 and 3) finite element computer program (Reference 6.15) for the feedwater nozzle (non-beltline) stresses. This analysis was performed to determine through-wall thermal and pressure stress distributions for the NMP1 feedwater nozzles due to a step-change thermal transient (Reference 6.16). The ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC Generic Letter 83-11, Supplement 1 (Reference 6.17), 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 following inputs were used in the finite element analysis:

With respect to operating conditions, stress distributions were developed for a thermal shock of 450°F, which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions. The stress results for a 450'F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting feedwaternozzle, as a shock of 450"F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 100lF/hr startup transient. Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 100F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region.

Heat transfer coefficients were calculated from the governing design basis stress report for the NMP1 feedwater nozzle and from a model of the heat transfer coefficient as a function of flow rate, as shown in Table 5 (Reference 6.16). The heat transfer coefficients were evaluated at flow rates that bound the actual operating conditions in the feedwater nozzles at NMP1.

A two-dimensional, axisymmetric finite element model of the feedwater nozzle was constructed (Figure 4) using the same modeling techniques that were employed to evaluate the feedwater nozzle in the governing design basis report. In order to properly model the feedwater nozzle, the analysis was performed as a penetration in a sphere and not in a cylinder. To make up for this difference in geometry, a conversion factor of Page 6 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report 3.2 times the cylinder radius was used to model the sphere (Reference 6.16). Material properties were evaluated at 325°F (Table 6) to conservatively bound the 100'F condition where the maximum stress occurred.

6.0 REFERENCES

1. Structural Integrity Associates Report No. SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007.

2. Structural Integrity Associates Calculation No. 0800297.301, Revision 1, "Revised Pressure-Temperature Curves," January 2009.

3. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

4. "Neutron Transport Analysis for Nine Mile Point Unit 1," Report Number MPM-405778, MPM Technologies, May 2006.

5. NRC Letter to NMPNS dated October 27, 2003, "Nine Mile Point Nuclear Station, Unit No. 1, Issuance of Amendment Re: Pressure-Temperature Limit Curves (TAC No.

MB6687)."

6.

NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

7. Structural Integrity Associates Calculation No. 0800297.300, Revision 1, "Evaluation of Adjusted Reference Temperature Shifts," August 2008.

8. NRC Generic Letter 92-01, Revision 1, "Reactor Vessel Structural Integrity," March 6, 1992.

9. NRC Generic Letter 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity," May 19, 1995.

10. NRC Letter to Niagara Mohawk Power Corporation (NMPC) dated March 30, 1994, "Generic Letter (GL) 92-01, Revision 1, 'Reactor Vessel Structural Integrity,' Nine Mile Point Nuclear Station Unit No. 1 (NMP-1) (TAC No. M83486)."

11. NRC Letter to NMPC dated August 26, 1996, "Closeout for Niagara Mohawk Power Corporation (NMPC) Response to Generic Letter 92-01, Revision 1, Supplement 1 for the Nine Mile Point Nuclear Station, Unit Nos. 1 & 2 (TAC Nos. M92700 and M927001)."

12. NRC Letter to NMPC dated June 25, 1999, "Response to Request for Additional Information Regarding Generic Letter 92-01, Revision 1, Supplement 1, 'Reactor Vessel Structural Integrity,' Nine Mile Point Nuclear Station, Unit Nos. 1 & 2 (TAC Nos. MA1200 and MA1201)."

Page 7 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report

13. NUREG-1900, "Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2," September 2006.

14. U. S. Nuclear Regulatory Commission, "Reactor Vessel Integrity Database Version 2.0.1," September 7, 2000.

15. ANSYS/Mechanical Release 6.1 (w/Service Packs 2 and 3), ANSYS, Inc., April 2002.

16. Structural Integrity Associates Calculation No. NMP-09Q-302, Revision 0, "Feedwater Nozzle Green's Functions for Nine Mile Point Unit 1."

17. NRC Generic Letter 83-11, Supplement 1, "Licensee Qualification for Performing Safety Analyses," June 24, 1999.

18. NRC Letter to NMPNS dated November 8, 2004, "Nine Mile Point Nuclear Station Unit Nos. 1 and 2 - Issuance of Amendments Re: Implementation of the Reactor Pressure Vessel Integrated Surveillance Program (TAC Nos. MC1758 and MC1759)."

19. G.E. Drawing No. 237E434, "Loadings Reactor Vessel."

20. "Neutron Transport Analysis for Nine Mile Point Unit 1," Report Number MPM-1209877, MPM Technologies, December 2009.

21. Engineering Change Notice, ECN No. N1-09-022 0800297.300-01.00 Rev. 000.

22. Calculation Change Notice, CCN No. N1-09-022 0800297.301-01.00 Rev. 000 Page 8 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report Figure 1: NMP1 Pressure Test (Curve A)

Curve is Valid Until End of Operating Cycle 22 1,900 1,800 1,700 1,600 1,500

• 2-*1,400 1,300 UJ U) 0/)

w 1,200 w

01,100 I-,

w 1,000 z

_ 900 800 w

700 U)

LU w

n 600 a-500 400 300 Beltline Region

- Bottom Head

............ Upper Vessel 200 Bolt-up Temp:

70'F 100 0

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL COOLANT TEMPERATURE (OF)

(MAINTAIN PLANT OPERATION TO THE RIGHT OF ALL CURVES)

Page 9 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report Figure 2: NMP1 Normal Operation (Heatup and Cooldown) -

Core Not Critical (Curve B)

Curve is Valid Until End of Operating Cycle 22 1,900 1,800 1,700 1,600 1,500

.'M1,400 0.

-J 1,300 U)

C,)

w1,200 01,100 l1,000 z

900 S800 LU 700 U)

(0 LU w

W 600 CL 500 I i i I I i Beltline Region

- Bottom Head

.......... Upper Vessel 7-7-

400 300 200 100 0-I.

I r

Bolt-up Temp:

70TF I

I I

In I I I

0 20 40 60 80100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL COOLANT TEMPERATURE (fF)

(MAINTAIN PLANT OPERATION TO THE RIGHT OF ALL CURVES)

Page 10 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report Figure 3: NMP1 Normal Operation (Heatup and Cooldown) -

Core Critical (Curve C)

Curve is Valid Until End of Operating Cycle 22 1,900 1,800 1,700 1,600 1,500

.'-i1,400 0.

- 1,300 w

U)

U, LU 1,200 o 1,100

<C.

w 1,000 900 I.-

-I 800 w

700 U) w W 600 0.

500 400 300 200 F-Minimum Criticality: 1 00F 100 0

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL COOLANT TEMPERATURE (TF)

MAINTAIN PLANT OPERATION TO THE RIGHT OF THE CURVE Page 11 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report Figure 4: NMPI Feedwater Nozzle Finite Element Model MN F~edwater Nozzle Finite Element Model Page 12 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report Table 1: NMPI Pressure Test (Curve A) - Beltline Region, EOC 22 Plant Component Vessel thickness, t Vessel Radius, R ART Kit Safety Factor Mý Temperature Adjustment Pressure Adjustment Pressure Adjustment nches iches

-=====>

31 no thermal effects) 6 EFPY F (instrument uncertainty) psig (hydrostatic pressure head for a full vessel at 70'F)

)sig (instrument uncertainty)

(1)

11)

Adjusted Temperature Pressure for Gauge for P-T Curve P-T Curve Gauge Fluid Temperature

(*F) 56 56 58 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 (ksi-Vin) 35.84 35.84 35.95 36.06 36.18 36.30 36.43 36.56 36.70 36.84 36.99 37.14 37.30 37.47 37.64 37.83 38.02 38.21 38.42 38.63 38.85 39.08 39.32 39.57 39.83 40.10 40.38 40.68 40.98 41.30 41.63 41.97 42.33 42.70 43.09 43.50 43.92 44.35 44.81 45.28 45.78 46.29 46.82 47.38 (ksi-Vin) 23.90 23.90 23.97 24.04 24.12 24.20 24.29 24.37 24.46 24.56 24.66 24.76 24.87 24.98 25.10 25.22 25.34 25.47 25.61 25.75 25.90 26.05 26.21 26.38 26.55 26.73 26.92 27.12 27.32 27.53 27.75 27.98 28.22 28.47 28.73 29.00 29.28 29.57 29.87 30.19 30.52 30.86 31.22 31.59 Pressure (psig) 0 647 649 651 653 655 657 660 662 665 667 670 673 676 679 683 686 690 693 697 701 705 710 714 719 724 729 734 739 745 751 757 764 771 778 785 792 800 809 817 826 835 845 855

(°F) 70 70 70 70 70 70 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120.

122 124 126 128 130 132 134 136 138 140 142 144 (psig) 0 609 611 613 615 617 620 622 625 627 630 633 635 638 642 645 648 652 656 659 663 668 672 676 681 686 691 696 702 708 714 720 726 733 740 747 755 763 771 779 788

,798 807 817 Page 13 of 22 PTLR-1 Revision 01.00

NMPI Pressure and Temperature Limits Report Table I (Continued)

Gauge Fluid Temperature

(°F) 142 144 146 148 150 152 154 156 158 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 224 K,ý (ksi-Vin) 47.96 48.56 49.19 49.84 50.52 51.23 51.96 52.73 53.52 54.35 55.22 56.11 57.05 58.02 59.04 60.09 61.19 62.33 63.52 64.76 66.04 67.38 68.78 70.23 71.74 73.32 74.95 76.66 78.43 80.28 82.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 K1.

(ksi-Vin) 31.97 32.37 32.79 33.23 33.68 34.15 34.64 35.15 35.68 36.24 36.81 37.41 38.03 38.68 39.36 40.06 40.79 41.55 42.35 43.17 44.03 44.92 45.85 46.82 47.83 48.88 49.97 51.10 52.29 53.52 54.80 56.13 57.52 58.96 60.47 62.03 63.66 65.35 67.12 68.95 70.86 72.85 Gauge Pressure (psig) 865 876 888 899 912 924 938 951 966 981 996 1013 1029 1047 1065 1084 1104 1125 1146 1168 1192 1216 1241 1267 1295 1323 1352 1383 1415 1449 1483 1519 1557 1596 1637 1679 1723 1769 1817 1866 1918 1972 (1)

Temperature for P-T Curve

.(°F) 146 148 150 152 154 156 158 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 224 226 228 (1)

Adjusted Pressure for P-T Curve (psig) 828 839 850 862 874 887 900 914 928 943 959 975 992 1009 1028 1047 1066 1087 1108 1131 1154 1178 1203 1230 1257 1285 1315 1346 1378 1411 1446 1482 1519 1558 1599 1641 1685 1731 1779 1829 1880 1934 (1) DATA IN THESE COLUMNS WERE USED TO PLOT P-T CURVES AND INCLUDE INSTRUMENT UNCERTAINTIES/STATIC HEAD CORRECTION.

Page 14 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report Table 2: NMP1 Normal Operation - Core Not Critical (Curve B), Beltline Region, EOC 22 Plant =

Component =

Vessel thickness, t =

Vessel Radius, R ART =

inches inches F====>

ksi??in 36 EFPY Kt =

Safety Factor =

Mm=

Temperature Adjustment =

°F (instrument uncertainty)

Pressure Adjustment Pressure Adjustment Heat Up and Cool Down Rate psig (hydrostatic pressure head for a full vessel at 707) psig (instrument uncertainty)

°F/Hr Gauge Fluid Temperature (F) 48 48 50 52 54 56 58 60 62 64 66 68 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 (1)

Temperature Gauge for P-T Curve Pressure (psig) l°F)

(1)

Adjusted Pressure for P-T Curve (osie)

K,, (ksi-Vin) 35.45 35.45 35.54 35.64 35.74 35.84 35.95 36.06 36.18 36.30 36.43 36.56 36.70 36.84 36.99 37.14 37.30 37.47 37.64 37.83 38.02 38.21 38.42 38.63 38.85 39.08 39.32 39.57 39.83 40.10 40.38 40.68 40.98 41.30 41.63 41.97 42.33 42.70 43.09 43.50 43.92 44.35 44.81 45.28 45.78 46.29 46.82 47.38 I

Klm (ksi-Vin) 11.27 11.27 11.31 11.36 11.41 11.46 11.52 11.57 11.63 11.69 11.76 11.82 11.89 11.96 12.04 12.11 12.19 12.28 12.37 12.46 12.55 12.65 12.75 12.86 12.97 13.08 13.20 13.33 13.46 13.59 13.73 13.88 14.03 14.19 14.36 14.53 14.71 14.90 15.09 15.29 15.50 15.72 15.95 16.18 16.43 16.69 16.95 17.23 305 305 306 308 309 310 312 313 315 317 318 320 322 324 326 328 330 332 335 337 340 342 345 348 351 354 357 361 364 368 372 376 380 384 389 393 398 403 408 414 420 425 432 438 445 452 459 466 70 70 70 70 70 70 70 72 74 76 78 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 142 144 146 148 150 152 0

225 226 228 229 230 232 233 235 237 238 240 242 244 246 248 250 252 255 257 260 263 265 268 271 274 278 281 284 288 292 296 300 304 309 313 318 323 329 334 340 346 352 358 365 372 379 387 Page 15 of 22 PTLR-1 Revision 01.00

I A NMPI Pressure and Temperature Limits Report Table 2 (Continued)

Gauge Fluid Temperature

(°F) 142 144 146 148 150 152 154 156 158 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 224 226 228 230 232 234 236 238 240 242 244 246 248 250 K,* (ksi-Vin) 47ý96 48.56 49.19 49.84 50.52 51.23 51.96 52.73 53.52 54.35 55.22 56.11 57.05 58.02 59.04 60.09 61.19 62.33 63.52 64.76 66.04 67.38 68.78 70.23 71.74 73.32 74.95 76.66 78.43 80.28 82.20 84.20 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 112.38 115.62 118.98 122.48 126.12 129.92 133.86 137.97 142.25 146.70 151.33 156.15 161.17 Kim (ksi-Vin) 17.52 17.82 18.14 18.46 18.80 19.16 19.52 19.91 20.30 20.72 21.15 21.60 22.07 22.55 23.06 23.59 24.14 24.71 25.30 25.92 26.57 27.24 27.93 28.66 29.41 30.20 31.02 31.87 32.76 33.68 34.64 35.64 36.68 37.77 38.89 40.07 41.29 42.56 43.88 45.26 46.69 48.18 49.74 51.35 53.03 54.78 56.60 58.50 60.47 62.53 64.67 66.89 69.21 71.62 74.13 Gauge Pressure (psig) 474 482 491 500 509 518 528 539 550 561 572 585 597 610 624 638 653 669 685 702 719 737 756 776 796 817 840 863 887 912 938 965 993 1022 1053 1084 1118 1152 1188 1225 1264 1304 1346 1390 1435 1483 1532 1583 1637 1692 1750 1811 1873 1938 2006 (1)

Temperature for P-T Curve

('F) 154 156 158 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 224 226 228 230 232 234 236 238 240 242 244 246 248 250 252 254 256 258 260 262 (1)

Adjusted Pressure for P-T Curve (psig) 394 403 411 420 429 439 449 459 470 481 493 505 517 531 544 559 573 589 605 622 639 657 676 696 716 738 760 783 807 832 858 885 913 942 973 1005 1038 1072 1108 1145 1184 1224 1266 1310 1356 1403 1452 1504 1557 1613 1670 1731 1793 1859 1926 (1)

DATA IN THESE COLUMNS WERE USED TO PLOT P-T CURVES AND INCLUDE INSTRUMENT UNCERTAINTIES/STATIC HEAD CORRECTION.

Page 16 of 22 PTLR-1 Revision 01.00

I..

NMP1 Pressure and Temperature Limits Report Table 3: NMP1 Normal Operation - Core Critical (Curve C), EOC 22 Plant =

Curve A Leak Test Temperature Curve A Leak Test Pressure Unit Pressure =

Flange RTNDT =

Adjusted P-T Curve Temperature

('F) 100 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 142 144 146 148 150 152 154 156 158 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 190 psig psig (hydrostatic pressure)

Adjusted P-T Curve Pressure (psig) 0 113 119 122 126 131 136 141 147 153 159 165 172 179 186 194 202 210 219 228 238 248 259 268 271 274 278 281 284 288 292 296 300 304 309 313 318 323 329 334 340 346 352 358 365 372 375 Page 17 of 22 PTLR-1 Revision 01.00

4,.

NMP1 Pressure and Temperature Limits Report Table 3 (Continued)

Adjusted P-T CurveAdjusted P-T CurveTempeatureCurve Temperature Pressure (psig)

(%F 192 375 194 375 196 375 198 375 200 375 200 420 202 429 204 439 206 449 208 459 210 470 212 481 214 493 216 505 218 517 220 531 222 544 224 559 226 573 228 589 230 605 232 622 234 639 236 657 238 676 240 696 242 716 244 738 246 760 248 783 250 807 252 832 254 858 256 885 258 913 260 942 262 973 264 1005 266 1038 268 1072 270 1108 272 1145 274 1184 276 1224 278 1266 280 1310 282 1356 284 1403 286 1452 288 1504 290 1557 292 1613 294 1670 296 1731 298 1793 300 1859 302 1926 Page 18 of 22 PTLR-1 Revision 01.00

NMP1 Pressure and Temperature Limits Report Table 4: NMP1 ART Calculations for 36 EFPY U,ci, 4-,

0~

Upper Shell Plate Upper Shell Plate Upper Shell Plate Lower Shell Plate Lower Shell Plate G-307-3 G-307-4 G-307-10 G-8-1 P2074 P2076 P2091 P2112 P'I *fA 28 40 20 36

-3 0.2 0.27 0.22 0.236 0.176 0.48 0.53 0.51 0.503 0.586 134.6 173.85 148.85 228.35 146.8 63.5 82.0 70.2 89.0 57.2 0

0 0

0 0

17 17 17 17 8.5 125.5 156.0 124.2 159.0 71.2 I

G-8-3/4 P2130A

-3 0,176 0.586 I

Upper Shell Axial Welds Lower Shell Axial Welds Circumferential Weld Seam 2-564A/C 2-564D/F 3-564 86054B 86054B 1248 4E5F 4E5F 4M2F

-50

-50 0.214 0.214 f) 1A 0.046 0.046 n

A7r 97.59 97.59 00 a 45.5 37.9 Q52 Q 0

0 n

22.8 19.0 1q C; 41.1 25.8

?7 q 1,

Upper Shell Plate Upper Shell Plate Upper Shell Plate Lower Shell Plate Lower Shell Plate G-307-3 G-307-4 G-307-10 G-8-1 G-8-3/4 7.125 7.125 7.125 7.125 7.125 1.781 1.781 1.781 1.781 1.781 2.OOE+18 2.OOE+18 2.OOE+18 1.34E+18 1.34E+18 0.652 0.652 0.652 0.652 0.652 1.30E+18 1.30E+18 1.30E+18 8.71E+17 8.71E+17 0.472 0.472 0.472 0.390 0.390 Upper Shell Axial Welds 2-564A/C 7.125 1.781 1.95E+18 0.652 1.27E+18 0.467 Lower Shell Axial Welds 2-564D/F 7.125 1.781 1.33E+18 0.652 8.65E+17 0.388 Circumferential Weld Seam 3-564 7.125 1.781 1.34E+18 0.652 8.71E+17 0.390 Page 19 of 22 PTLR-1 Revision 01.00

-4 NMPI Pressure and Temperature Limits Report Table 5: Heat Transfer Coefficients for NMP1 Feedwater Nozzle 0% Flow Case Heat Transfer Temperature Coefficient Region

('F)

(Btu/hr-ft2-°F) 1 550.0 205.1 2

550.0 205.1 3

550.0 205.1 4

550.0 205.1 100% Flow Case Heat Transfer Temperature Coefficient Region

('F)

(Btu/hr-ft 2 -°F) 1 100.0 2108.8 2

325.0 673.9 3

325.0 191.8 4

550.0 1000.0 Page 20 of 22 PTLR-1 Revision 01.00

p

-4 NMP1 Pressure and Temperature Limits Report Table 6: Feedwater Nozzle Material Properties Material Properties Al Steels: Poissonrs Ratio Density 0,3 0.283 T

F 300 350 400 325 T

F 300 350 400 325 T

F 300 350 400 325 a

in~nt 7.74E 7 88E

&.01E 7.81E Reactor Vessel Plate (SA 302 Gr.B) [5, Material Group D]

E Thermal Concductivity, K Thermal Diffutsidy Specific Heal, Cp xF psi BTU/hr<tt*F ft I/hr B TU/b *F

-06 2.80E+07 24.7 0.42 0.12

-06 24.7 0,409 0.123

-06 2-74E"07 24.6 0.398 0.126

-00 2,79E+07 243 0.4145 0.1215 Nozzle Forging (SA 336 with Code Case 1236-1) [5, Material Group A]

E Tihermal Conductivity, K Thermal Difftusivily Specific Heal, Cp iin/in"F psi BTUL/hrwfIF ft Z#A2r B TUJ/b XF 7.30E-06 2,85E+07 23.9 0.406 0.120 7.49E-06 23.7 0.396 0.122 7GCE-06 2.79E+07 23.6 0.385 0.125 7.395E-06 '2-84E+07 23.8 0.401 0.121 Safe End (CS4 SA-105 Gr. II) [5, Material. Group 2J E

F Thermal Conductivity, K Thermal Diffusivity Specific H-eat in/in"F psi BTU/hr'UrF ftR 'hr B TUIb *tF 7,18E-06 2.81E407 28.4 0.481 0.1207 7.47E-06 28.0 0.464 0.1234 Co 2.75E+07 7.325E-06 r2.80E+07 28.2 0.4725 0.1221 Page 21 of 22 PTLR-1 Revision 01.00

0 4,

NMP1 Pressure and Temperature Limits Report APPENDIX A NMP1 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM NMP1 has replaced the original materials surveillance program with the BWRVIP Integrated Surveillance Program (ISP). This program meets the requirements of 10 CFR 50, Appendix H, for integrated surveillance programs, and has been approved by the NRC (see NMP1 License Amendment No. 184, Reference 6.18). The representative plate material from the ISP is not the same heat number as the target plate in the NMP1 vessel. Also, the representative weld material is not the same heat number as the target weld in the NMP1 vessel. However, there is one matching plate heat number (heat number P2130-2) in the Supplemental Surveillance Program (SSP). Irradiated data is available from SSP capsules A, B, D, G, E, and I (Reference 6.7). Under the ISP, there is one weld heat that is scheduled to be tested in 2017.

Representative surveillance capsule materials for the NMP1 weld are contained in the Hatch Unit 2 surveillance capsule program. Under the Supplemental Surveillance Program (SSP),

there are no additional representative capsule materials to be tested.

Page 22 of 22 PTLR-1 Revision 01.00