L-2010-259, Extended Power Uprate Licensing Report, Attachment 5; Appendix G, WCAP-17197-NP, Revision 0, RCS Pressure and Temperature Limits and Low-Temperature Overpressure Protection Report for 54 Effective Full Power Years
| ML103560511 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/28/2010 |
| From: | Byrne S, Fournier R, Ganta B Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-2010-259 WCAP-17197-NP, Rev 0 | |
| Download: ML103560511 (53) | |
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{{#Wiki_filter:St. Lucie Unit 1 L-2010-259 Docket No. 50-335 Attachment 5 St. Lucie Unit 1 Extended Power Uprate Licensing Report Attachment 5 Appendix G WCAP-17197-NP Revision 0 St. Lucie Unit 1 RCS Pressure and Temperature Limits and Low-Temperature Overpressure Protection Report For 54 Effective Full Power Years This coversheet plus 52 pages St. Lucie Unit 1 EPU Licensing Report App. G-1 WCAP-17197-NP Revision 0
Westinghouse Non-Proprietary Class 3 WCAP-17197-NP February 2010 Revision 0 St. Lucie Unit 1 ReS Pressure and Temperature Limits and Low-Tem peratu re Overpressure Protection Report for 54 Effective Full Power Years
- Westinghouse
LEGAL NOTICE This report is an account of work performed by Westinghouse Electric Company LLC for Florida Power and Light Company. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf: A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report. WCAP-17197-NP February 2010 Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17197-NP Revision 0 St. Lucie Unit 1 RCS Pressure and Temperature Limits and Low-temperature Overpressure Protection Report for 54 Effective Full Power Years R. S. Fournier* Systems & Equipment Engineering II B. Reddy Ganta* Major Reactor Component Design & Analysis - I S. T. Byrne* Reactor Internals Design & Analysis II February 2010 Reviewer: M. J. Thibodeau* Systems & Equipment Engineering II Reviewer: G. Z. Hall* Major Reactor Component Design & Analysis - I Reviewer: P. R. Sotherland* Reactor Internals Design & Analysis II Approved: M. Gancarz*, Manager Systems & Equipment Engineering II
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Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355
© 2010 Westinghouse Electric Company LLC All Rights Reserved WCAP-17197-NP_14.doc-022610
ii COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. Information in this report is the property of and contains copyright material owned by Westinghouse Electric Company LLC and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document and the material contained therein in strict accordance with the terms and conditions of the agreement under which it was provided to you. As a sponsor of this task, you are permitted to make the number of copies of the information contained in this report that are necessary for your internal use in connection with your implementation of the report results for your plant(s) in your normal conduct of business. Should implementation of this report involve a third party, you are permitted to make the number of copies of the information contained in this report that are necessary for the third party's use in supporting your implementation at your plant(s) in your normal conduct of business if you have received the prior, written consent of Westinghouse Electric Company LLC to transmit this information to a third party or parties. All copies made by you must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary. The NRC is permitted to make the number of copies beyond those necessary for its internal use that are necessary in order to have one copy available for public viewing in the appropriate docket files in the NRC public document room in Washington, DC if the number of copies submitted is insufficient for this purpose, subject to the applicable federal regulations regarding restrictions on public disclosure to the extent such information has been identified as proprietary. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary. WCAP-17197-NP February 2010 Revision 0
iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES ...................................................................................................................................... v 1 INTRODUCTION ........................................................................................................................1-1 2 PRESSURE-TEMPERATURE LIMITS......................................................................................2-1 2.1 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS.....................................2-1 2.2 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS ............................................................................................................................2-4 2.2.1 Application of Pressure Correction Factors.....................................................2-6 2.3 THERMAL ANALYSIS METHODOLOGY ................................................................2-10 2.4 COOLDOWN LIMIT ANALYSIS ................................................................................2-10 2.5 HEATUP LIMIT ANALYSIS ........................................................................................2-14 2.6 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS .........................2-16 2.7 LOWEST SERVICE TEMPERATURE, MINIMUM BOLTUP TEMPERATURE, FLANGE LIMIT TEMPERATURE, MINIMUM PRESSURE LIMITS AND LTOP ENABLE TEMPERATURES ........................................................................................2-18 2.8 DATA .............................................................................................................................2-20 3 LOW-TEMPERATURE OVERPRESSURE PROTECTION ......................................................3-1 3.1 GENERAL.......................................................................................................................3-1 3.2 METHOD AND ASSUMPTIONS ..................................................................................3-1 3.3 PRESSURE TRANSIENT ANALYSES .........................................................................3-2 3.3.1 Energy Addition Transients .............................................................................3-2 3.3.2 Mass Addition Transients ................................................................................3-7 3.3.3 Controlling Pressures.....................................................................................3-17 3.4 LIMITING CONDITIONS FOR OPERATION ............................................................3-17 3.5
SUMMARY
OF PROPOSED CHANGES....................................................................3-18 4 REFERENCES .............................................................................................................................4-1 APPENDIX A TECHNICAL SPECIFICATION FIGURES .................................................................. A-1 WCAP-17197-NP February 2010 Revision 0
iv LIST OF TABLES Table 2-1 St. Lucie Unit 1 Reactor Vessel Beltline Materials..........................................................2-2 Table 2-2 St. Lucie Unit 1 Reactor Vessel Beltline Material Data for 1/4T and 3/4T .....................2-3 Table 2-3 St. Lucie Unit 1 Controlling Materials and their ARTs....................................................2-4 Table 2-4 Cooldown Allowable Pressures, Uncorrected................................................................2-12 Table 2-5 St. Lucie Unit 1 Cooldown and Heatup Allowable Pressure 54 EFPY, Adjusted to Actual PZR Pressure, APCF......................................................................................................2-13 Table 2-6 St. Lucie Unit 1 Cooldown and Heatup Allowable Pressure 54 EFPY, Adjusted to Indicated PZR Pressure, IPCF .......................................................................................2-14 Table 2-7 Heatup Allowable Pressures, Uncorrected.....................................................................2-15 Table 2-8 St. Lucie Unit 1 Hydrostatic Test P-T Limit Data..........................................................2-17 Table 2-9 LTOP Enable Temperature Limits .................................................................................2-19 Table 3-1 Maximum Transient Pressures at 530 psia Setpoint ........................................................3-4 Table 3-2 Maximum Transient Pressures at 350 psia Setpoint* ......................................................3-4 Table 3-3 LTOP Requirements, 54 EFPY ......................................................................................3-17 WCAP-17197-NP February 2010 Revision 0
v LIST OF FIGURES Figure 2-1 St. Lucie Unit 1 Cooldown P-T Limits 54 EFPY, APCF Adjusted to PZR Pressure ..............2-8 Figure 2-2 St. Lucie Unit 1 Heatup P-T Limits 54 EFPY, APCF Adjusted to PZR Pressure ...................2-8 Figure 2-3 St. Lucie Unit 1 Cooldown P-T Limits 54 EFPY, IPCF Adjusted to Indicated PZR Pressure............................................................................................................................2-9 Figure 2-4 St. Lucie Unit 1 Heatup P-T Limits 54 EFPY, IPCF Adjusted to Indicated PZR Pressure............................................................................................................................2-9 Figure 3-1 St. Lucie 1, Energy Addition Transient Case 1, PORV, PSET = 350 psia, TC = 140°F......3-5 Figure 3-2 St. Lucie 1, Energy Addition Transient Case 2, PORV, PSET = 350 psia, TC = 200°F .............3-5 Figure 3-3 St. Lucie 1, Energy Addition Transient Case 3, PORV, PSET = 530 psia, TC = 200°F....3-6 Figure 3-4 St. Lucie 1, Energy Addition Transient Case 4, PORV, PSET = 530 psia, TC = 300°F .............3-6 Figure 3-5 St. Lucie Unit 1 LTOP Mass Addition Transient Case 1 2 HPSI + 3 CPs, 300°F...........3-9 Figure 3-6 St. Lucie Unit 1 LTOP Mass Addition Transient Case 3 2 HPSI + 3 CPs, 220°F..........3-10 Figure 3-7 St. Lucie Unit 1 LTOP Mass Addition Transient Case 4 1 HPSI + 3 CPs, 300°F.......... 3-11 Figure 3-8 St. Lucie Unit 1 LTOP Mass Addition Transient Case 6 1 HPSI + 3 CPs, 220°F..........3-12 Figure 3-9 St. Lucie Unit 1 LTOP Mass Addition Transient Case 7 3 CPs, 220°F .........................3-13 Figure 3-10 St. Lucie Unit 1 LTOP Mass Addition Transient Case 8 3 CPs, 140°F .........................3-14 Figure 3-11 St. Lucie Unit 1 LTOP Mass Addition Transient Case 9 Single HPSI Pump, 220°F.....3-15 Figure 3-12 St. Lucie Unit 1 LTOP Mass Addition Transient Case 10 Single HPSI Pump, 140°F...3-16 WCAP-17197-NP February 2010 Revision 0
1-1 1 INTRODUCTION The following sections describe the basis for developing reactor vessel beltline pressure-temperature (P-T) limitations and low-temperature overpressure protection (LTOP) requirements for the St. Lucie, Unit 1, Nuclear Generating Station. These limits are calculated to meet the regulations of the U.S. Nuclear Regulatory Commission (NRC) 10 CFR 50, Appendix A (Reference 1), Design Criterion 14 and Design Criterion 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected and tested in order to have an extremely low probability of abnormal leakage, of rapid propagating failure, and of gross rupture. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance and testing the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. The P-T limits are developed using the requirements of 10 CFR 50 Appendix G (Reference 2). This appendix describes the requirements for developing the P-T limits and provides the general basis for these limitations. The margins of safety against fracture provided by the P-T limits using the requirements of 10 CFR 50 Appendix G are equivalent to those recommended in the ASME Boiler and Pressure Vessel Code Section III, Appendix G, Fracture Toughness Criteria for Protection against Failure (Reference 3). The general guidance provided in those procedures was utilized to develop the St. Lucie Unit 1 P-T limits with the requisite margins of safety for the heatup and cooldown conditions. The reactor pressure vessel beltline P-T limits are based upon the irradiation damage prediction methods of Regulatory Guide 1.99, Revision 2 (Reference 4). This methodology was used to calculate the limiting material adjusted reference temperatures for St. Lucie Unit 1 utilizing fluence values corresponding to 54 effective full power years (EFPY). This report provides reactor vessel beltline P-T limits generated in accordance with 10 CFR 50, Appendix G for 54 EFPY. The events analyzed are the isothermal, 20 through 100°F/hr cooldown conditions and both 50°F/hr and 70°F/hr heatup conditions. These conditions were analyzed to provide a data base of reactor vessel P-T limits for use in establishing LTOP requirements. LTOP requirements were established based upon the guidance provided in U.S. NRC Standard Review Plan (SRP), 5.2.2 (Reference 5). Using this guidance, the limiting transient pressures were determined for mass and energy addition transients to establish the appropriate LTOP setpoints, heatup and cooldown rates, and administrative requirements. Based upon the P-T limit analyses and LTOP requirements provided within this report, no limiting vessel operability issues are anticipated. WCAP-17197-NP February 2010 Revision 0
2-1 2 PRESSURE-TEMPERATURE LIMITS 2.1 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS In order to develop P-T limits over the design life of the reactor vessel, adjusted reference temperatures (ARTs) for the controlling beltline material need to be determined. The ARTs for the St. Lucie Unit 1 reactor vessel beltline materials were calculated for 54 EFPY at both the 1/4t and 3/4t locations. The vessel material with the highest ART (i.e., the controlling material) was used as the input to the P-T limits for St. Lucie Unit 1. The ART values have been calculated using the procedures in Regulatory Positions 1.1 and 1.2 of Regulatory Guide 1.99, Revision 2 (Reference 4). The calculation for the ART values for each material in the beltline is shown below. ART = Initial RTNDT + RTNDT + margin The neutron fluence is attenuated through the vessel wall using the nominal reactor vessel thickness of 8.625 inches (Reference 6), conservatively neglecting cladding thickness. The material input data are listed in Table 2-1. The St. Lucie Unit 1 reactor vessel is weld limited, with the limiting 1/4t and 3/4t ART values of 210°F and 156°F, respectively. The 54 EFPY ART projections for all beltline materials are provided in Table 2-2. The following information provides the basis for the calculated ART values for St. Lucie Unit 1.
- 1. The material data were obtained from References 7, 8, and 9, including copper content, nickel content and initial reference temperature (initial RTNDT). These data are summarized in Table 2-1 for St. Lucie Unit 1.
- 2. The peak neutron fluence at 54 EFPY for the Unit 1 beltline region was determined to be 4.21 x 1019 n/cm2 (E>1 MeV) for the base metal and the circumferential weld, and 2.74 x 1019 n/cm2 (E>1 MeV) for the axial welds. The fluence analysis was based on a plant and fuel-cycle-specific basis for the first 26 reactor operating cycles. For Cycles 1 through 23, the effective full power is 2700 MWt. Cycles 24 and 25 are EPU transition cycles at 3020 MWt and Cycle 26 is representative of the equilibrium for EPU at 3020 MWt. Projections were made to 54 EFPY beginning from the end of Cycle 24 assuming the uprated core power of 3020 MWt.
- 3. The reactor vessel beltline thickness was 8.625 inches. (Reference 6).
- 4. Calculations were based on the procedures in Regulatory Guide 1.99, Revision. 2 (Reference 4).
ARTs for all beltline materials at the 1/4t and 3/4t locations after 54 EFPY were calculated using Regulatory Guide 1.99, Revision 2. The results of the calculation are listed in Table 2-2 for St. Lucie Unit
- 1. The vessel material with the highest ART is shown in Table 2-3. These limiting ART values were then used to develop the P-T limits for the corresponding time period. In the case of St. Lucie Unit 1, the lower shell axial welds (3-203 A/C) are controlling at the 1/4t and 3/4t location after 54 EFPY based on the predicted ART values of 210°F and 156°F, respectively.
WCAP-17197-NP February 2010 Revision 0
2-2 Table 2-1 St. Lucie Unit 1 Reactor Vessel Beltline Materials Material Description Material Heat Number Cu (%) Ni (%) Initial RTNDT Intermediate Shell Plate C-7-1 A-4567-1 0.11 0.64 0°F Intermediate Shell Plate C-7-2 B-9427-1 0.11 0.64 -10°F Intermediate Shell Plate C-7-3 A-4567-2 0.11 0.58 10°F Lower Shell Plate C-8-1 C-5935-1 0.15 0.56 20°F Lower Shell Plate C-8-2 C-5935-2 0.15 0.57 20°F Lower Shell Plate C-8-3 C-5935-3 0.12 0.58 0°F Intermediate to Lower Shell Girth Weld 9-203 90136 0.27 0.07 -60°F Intermediate Shell Axial Weld 2-203 A/C A-8746 and 34B009 0.19 0.09 -56°F Lower Shell Axial Weld 3-203 A/C 305424 0.27 0.63 -60°F WCAP-17197-NP February 2010 Revision 0
2-3 Table 2-2 St. Lucie Unit 1 Reactor Vessel Beltline Material Data for 1I4T and 3/4T Beltline Initial ~RTNOT (1) Margin(2) CF (OF) FF al (OF) at. (OF) ART(3) (OF) Material RTNOT (OF) (OF) (OF) 1/4 T C-7-1 O°F 74.6 1.246997 93 0 17 34 127 C-7-2 -10°F 74.6 1.246997 93 0 17 34 117 C-7-3 10°F 73.8 1.246997 92 0 17 34 136 C-8-1 20°F 81.80(4) 1.246997 102 0 8.5 17 139 C-8-2 20°F 82.22(4) 1.246997 103 0 8.5 17 140 C-8-3 O°F 62.68(4) 1.246997 78 0 8.5 17 95 9-203 -60°F 84.97(4) 1.246997 106 0 14 28 74 2-203 A/C -56°F 90.7 1.135602 103 17 28 65.5 112 3-203 A/C -60°F 188.8 1.135602 214 0 28 56 210 3/4 T C-7-1 O°F 74.6 0.967610 72 0 17 34 106 C-7-2 -10°F 74.6 0.967610 72 0 17 34 96 C-7-3 10°F 73.8 0.967610 71 0 17 34 115 C-8-1 20°F 81.80(4) 0.967610 79 0 8.5 17 116 C-8-2 20°F 82.22(4) 0.967610 80 0 8.5 17 117 C-8-3 O°F 62.68(4) 0.967610 61 0 8.5 17 78 9-203 -60°F 84.97(4) 0.967610 82 0 14 28 50 2-203 A/C -56°F 90.7 0.848024 77 17 28 65.5 86 3-203 A/C -60°F 188.8 0.848024 160 0 28 56 156 Notes: (1 ) ~RT NOT = CF
- FF Margin = M = 2 OJ +0t.
2 2 (2) (3) ART = Initial RTNOT + ~RT NOT + M (4) Regulatory Position 2.1 [6] WCAP-17197-NP February 2010 Revision 0
2-4 Table 2-3 St. Lucie Unit 1 Controlling Materials and their ARTs Reactor Vessel Material ART at 54 EFPY, °F Location ID No. 1/4T 3/4T Lower Shell Axial Welds 3-203 A/C 210 156 According to Position 1.1 of Regulatory Guide 1.99, Revision 2 (Reference 4), there are two values of uncertainty. One is specific to the value of the initial RTNDT. If the RTNDT is derived in accordance with NB2300 of the ASME Boiler and Pressure Vessel Code, Section III, the uncertainty is assumed to be zero. If the RTNDT is based a generic value, then the uncertainty is derived from the data used to establish the generic value. For the one case in which a generic value was used, intermediate shell axial welds (2-203 A/C), the uncertainty associated with the -56°F initial RTNDT was 17°F. The other uncertainty applies to the shift prediction. If the prediction applies to base metal (plate), the one-sigma uncertainty was 17°F. If the prediction applies to weld metal, the one-sigma uncertainty was 28°F. In the case where the chemistry factor was based on credible surveillance data (Position 2.1 of Regulatory Guide 1.99), the one-sigma uncertainty can be reduced to 8.5°F for base metal, and 14°F for welds. 2.2 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS The analytical procedure for developing reactor vessel P-T limits utilizes the methods of linear elastic fracture mechanics (LEFM) and guidance found in the ASME Boiler and Pressure Vessel Code Section XI, Appendix G (Reference 3), in accordance with the requirements of 10 CFR 50 Appendix G. For these analyses, the Mode I (opening mode, according to fracture mechanics terminology) stress intensity factors were used for the solution basis. The St. Lucie, Unit 1, 54 EFPY P-T limits analysis utilizes a Westinghouse-developed and quality assured computer code to generate P-T limits for the reactor beltline region. That computer code uses superposition technique and influence coefficients to calculate these curves. The reactor coolant system (RCS) P-T limit curves were based on the beltline P-T limits for a set of heatup and cooldown rates. These curves were then adjusted to represent pressurizer (PZR) pressure conditions (the adjustment addresses both the RCS hydraulic pressure drop due to flow and PZR-to-beltline region elevation) and, where appropriate, adjusted for temperature and pressure instrumentation uncertainties. The final P-T limits include the minimum bolt-up temperature, lowest service temperature, and the flange limit. The minimum bolt-up temperature is specified in Reference 6. The LTOP enable temperatures were also determined using heat transfer results from the P-T limits analysis, and applying ASME Boiler and Pressure Vessel Code Section XI, Appendix G methodology. WCAP-17197-NP February 2010 Revision 0
2-5 The temperature distribution throughout the reactor vessel wall was characterized by a partial differential equation, defined for the applicable boundary conditions and geometry, and solved numerically. The numerical solution uses a finite element model to determine wall temperature as a function of radius, time, and thermal rate. This method utilizes three-noded, isoparametric finite elements suitable for one-dimensional, axisymmetric radial conduction-convection heat transfer. The wall was divided into 11 elements. The first element represents cladding, and the remaining 10 elements represent base metal. The analysis code utilizes convective boundary conditions on the inside wall and an insulation boundary on the outside wall of the reactor vessel. Variation of material properties through the wall was permitted, which allows for the change in material thermal properties between the cladding and the base metal. The reactor vessel beltline region was analyzed assuming a semi-elliptical surface flaw oriented in the axial direction, with a depth of one-quarter of the reactor vessel beltline thickness. This assumed flaw has an aspect ratio of one to six. The postulated flaw was analyzed at both the inside diameter location (referred to as the 1/4t location) and the outside diameter location (referred to as the 3/4t location) to ensure that the most limiting condition was achieved. At each of the postulated flaw locations, the Mode I stress intensity factor, KI, produced by each of the specified loadings, was calculated. The summation of the KI values was compared to reference stress intensity, KIC, which is the critical value of KI for the involved material and temperature. The result of this method is a relationship of pressure versus temperature for reactor vessel operating limits, which conservatively precludes brittle fracture. KIR is obtained from a reference fracture toughness curve for reactor vessel low alloy steels, and is defined in Appendices A and G of Section XI of the ASME Code (References 10 and 3 respectively). In this calculation, KIR was defined as KIC, and it was the lower bound of static initiation critical KI values measured as a function of temperature. This governing curve is defined by Equation 3 below. For operational events, P-T limits were calculated using the following equation. 2K IM K IT d K IC Equation 1 where: KIM = Membrane (pressure) stress intensity factor, ksi in KIT = Thermal stress intensity factor, ksi in KIC = Reference stress intensity factor, ksi in Rearranging the terms in the above equation: K IC K IT K IM Equation 2 2 WCAP-17197-NP February 2010 Revision 0
2-6 Allowable pressure was then computed using the allowable membrane stress intensity factor from Equation 2 and the pressure influence coefficients. The fracture toughness is shown in the following. K IC 33.2 20.734e 0.02 T RTNDT Equation 3 For the hydrostatic test limits, the structural factor 2 in Equation 1 is replaced by 1.5. 1.5K IM d K IC Equation 4 K IC K IM Equation 5 1 .5 For any instant during the postulated heatup or cooldown, KIC was calculated at the metal temperature and at the adjusted RTNDT at the tip of the flaw. The temperature distribution and the temperature at the flaw tip were calculated using a one-dimensional, three-noded isoparametric finite element suitable for one-dimensional radial conduction-convection heat transfer analysis. The fracture mechanics algorithms use a superposition technique using influence coefficients to calculate the Mode I stress intensity factors. At the conditions of 54 EFPY, isothermal and transient conditions were analyzed. The cooldown transients analyzed at rates of 10°F/hr, 20°F/hr, 30°F/hr, 40°F/hr, 50°F/hr, and 100°F/hr began at a bulk coolant temperature of 550°F and terminated at 80°F. The heatup transient analyzed had rates of 50°F/hr and 70°F/hr, and began at a bulk temperature of 80°F, terminating at 550°F. The hydrostatic limits were obtained for the isothermal condition only. 2.2.1 Application of Pressure Correction Factors The P-T limits, as directly calculated by ASME methodology, typically represent the limiting material conditions at the reactor vessel beltline. However, these beltline P-T limits could not be used directly by the plant operations staff, since pressure measurement in the RCS was limited to the PZR and, as such, the beltline values require adjustment to representative values relative to the PZR location. Pressure correction factors (PCFs) were used to adjust the beltline P-T limits to PZR pressure. These PCFs were updated for the current plant operations for this analysis, and consist of:
- 1. The pressure difference due to the static head of fluid between the PZR pressure instrument nozzle elevation and reactor vessel beltline lowest point
- 2. The flow-induced pressure drop between the applicable point in the reactor vessel and hot leg surge line nozzle, due to flow resulting from operating reactor coolant pumps (RCPs)
- 3. The uncertainty associated with the pressure instrumentation, as applicable WCAP-17197-NP February 2010 Revision 0
2-7 These PCFs were applied to the beltline P-T limits in two ways. An actual pressure correction factor (APCF) was applied to the beltline P-T limits to provide representative actual (or analysis) values relative to the PZR location. An APCF was developed from plant data associated with items (1) and (2) in the prior paragraphs explanation. APCFs have been developed to represent multiple plant operating conditions (e.g., combinations of operating RCPs). A bounding static head condition (1) and both a bounding consideration of three operating RCPs (2) as well as a bounding consideration of two operating RCPs (2) were selected. These two PCFs encompass the entire LTOP range for temperatures above 200°F, including three RCPs and temperatures below 200°F limited to two RCPs. These updated APCFs were developed to be bounding for each condition. For temperatures above 200°F, the APCF value was 72.8 psid; and for temperatures below 200°F, the APCF value was 59.8 psid. The potential condition of up to three operating RCPs fully bounds the plant operating conditions within the LTOP applicable range. Current plant procedures limit the operation of four RCPs to greater than 500°F. An inspection of the P-T limits, shown in Figures 2-1 and 2-2, indicates that the most limiting pressures were greater than 2400 psia at any temperature value above 300°F. Revisions of plant procedures will be established to ensure no more than two RCPs are operating below 200°F. Due to uncertainties in the PZR pressure instrument loop components, indicated PZR pressure observed by control room operators can differ from the actual PZR pressure. If unaccounted for, the actual PZR pressure can be greater than the indicated PZR pressure, which could potentially lead to a violation of the actual P-T limits. To prevent this, an indicated pressure correction factor (IPCF) was applied. For temperatures above 200°F the IPCF value was 107.8 psid, and for temperatures below 200°F the IPCF value was 94.8 psid. This accounts for the instrumentation uncertainty (item 3 from the previous page), in addition to the previously described adjustments for actual limits, to represent the indicated P-T limits. In conditions where the indicated P-T limits are developed (IPCF are applied), corresponding conservative temperature value adjustments are accommodated by a temperature correction factor, which acknowledges the possible uncertainty of the temperature indication loop. A value of 7°F was applied to adjust the P-T limits as well as the LTOP enable temperature, presented in the Technical Specification figures, since this represents the control room instrument error. Figures 2-3 and 2-4 show the limiting indicated pressures as a function of indicated temperature, accounting for this uncertainty. As a note, an additional 2°F must be applied to the setpoint for PORV actuation to account for the total loop uncertainty of 9°F associated with the LTOP actuation channels. WCAP-17197-NP February 2010 Revision 0
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, Temperature 275.2°F 50 100 150 200 250 300 350 400 Res Temperature (OF)
Figure 2-3 St. Lucie Unit 1 Cooldown P-T Limits 54 EFPY, IPCF Adjusted to Indicated PZR Pressure 2500
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o o 50 100 150 200 250 300 350 400 RCS Temperature (OF) Figure 2-4 St. Lucie Unit 1 Heatup P-T Limits 54 EFPY, IPCF Adjusted to Indicated PZR Pressure WCAP-17l97-NP February 2010 Revision 0
2-10 2.3 THERMAL ANALYSIS METHODOLOGY The thermal stress intensity factors were found by using the temperature differences through the wall as a function of transient time. They were then subtracted from the available KIR value to calculate the allowable pressure stress intensity factor and, consequently, the allowable pressure. Equation 1 provides the expression used to derive P-T limits. The superposition technique used was temperature profile-based rather than the commonly used stress profile-base. A third-order polynomial fit to the temperature distributions in the wall was used (Reference 11). x x x T( x ) C 0 C1 (1 ) C 2 (1 ) 2 C 3 (1 ) 3 Equation 6 h h h where: T(x) = temperature at radial location x from inside wall surface C0-C3 = coefficients in polynomial fit x = distance through beltline wall, inches h = beltline wall thickness, inches The unit KI values were calculated for each term of the polynomial using a two-dimensional finite element code. These unit values were used to determine the total KI value for the applied loads under any general temperature profile in the wall that develops during the thermal transient. The thermal stress intensity factor is represented by Equation 7. 3 K IT (a ) Ci K i Equation 7 i 0 where: KIT = Thermal stress intensity factor Ci = Coefficients in polynomial fit Ki = Polynomial influence coefficients Temperature-based influence coefficients for determining the thermal stress intensity factor, KIT were used. Using Reference 12 methods, these coefficients were computed using a two-dimensional reactor vessel model with a crack adjusted to account for three-dimensional effects. 2.4 COOLDOWN LIMIT ANALYSIS During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, KIM, and the thermal stress intensity factor, KIT, acting in unison to create high stress intensity. At the reactor vessel outside wall, the tensile pressure stress and the compressive thermal stress act in opposition, resulting in a lower total stress than at the WCAP-17197-NP February 2010 Revision 0
2-11 inside wall location. Also, neutron embrittlement, the shift in RTNDT, and the reduction in fracture toughness were less severe at the outside wall when compared to the inside wall location. Consequently, the inside flaw location is more limiting for the cooldown event. The reference stress intensities were determined by utilizing the material metal temperature and adjusted RTNDT at the 1/4t and 3/4t locations. The finite element method was used to perform the heat transfer analysis and the resulting through-wall temperature gradient, calculated for the assumed cooldown rate, is used to determine the thermal stress intensity factor. The thermal stress intensity factors were determined by using the temperature difference through the wall as a function of transient time. Those factors were then, subtracted from the available KIC value to calculate the allowable pressure stress intensity factor and, consequently, the allowable pressure. The cooldown P-T curves were thus generated by calculating the allowable pressure on the reference flaw at the 1/4t and 3/4t locations. This was based upon Equation 2 of Section 2.2. To develop a minimum P-T limit for the cooldown event, the isothermal P-T limit must be calculated, after which the isothermal P-T limit was compared to the P-T limit associated with a cooling rate. The more restrictive allowable P-T limit was chosen, which results in a minimum limit curve for the reactor vessel beltline. Table 2-4 shows the P-T limits results for conditions at the beltline (without applied correction factors) for cases for isothermal and 20°F/hr, 30°F/hr, 40°F/hr, 50°F/hr, and 100°F/hr cooldown. Tables 2-5 and 2-6 provide results that include the APCF and IPCF, respectively. APCF data were compared to the design basis LTOP transient results, which were also referenced to the PZR pressure location. IPCF data were used for the recommended Technical Specification P-T Limit figure changes. Uncorrected values are provided for completeness. WCAP-17197-NP February 2010 Revision 0
2-12 Table 2-4 Cooldown Allowable Pressures, Uncorrected Isothermal 20°F/hr 30°F/hr 40°F/hr 50°F/hr 100°F/hr Temperature Pall Pall Pall Pall Pall p. 1I (OF) (Dsia) IDsia) IDsia) IDsial (Dsia) IDSia) 80 657 577 537 498 459 271 90 663 584 546 507 469 287 100 671 594 556 518 481 306 110 680 605 568 532 496 330 120 692 619 583 548 514 359 130 706 636 602 569 536 394 140 723 656 624 594 564 437 150 744 681 652 624 597 490 160 769 712 686 660 637 554 170 801 750 727 706 687 632 180 839 796 777 761 748 724 190 885 851 838 828 821 817 199.9 942 919 913 910 910 910 200 942 920 914 911 911 911 210 1,012 1,003 1,003 1,003 1,003 1,003 220 1,097 1,097 1,097 1,097 1,097 1,097 230 1,200 1,200 1,200 1,200 1,200 1.200 240 1,327 1,327 1,327 1,327 1,327 1,327 250 1,481 1,481 1,481 1,481 1,481 1,481 260 1,670 1,670 1,670 1,670 1,670 1,670 270 1,901 1,901 1,901 1,901 1,90') 1,901 280 2,182 2,182 2,182 2,182 2,182 2,182 290 2,526 2,526 2,526 2,526 2,526 2,526 300 2,947 2,947 2,947 2,947 2,947 2.947 WCAP-17197-NP February 2010 Revision 0
2-13 Table 2-5 St. Lucie Unit 1 Cooldown and Heatup Allowable Pressure 54 EFPY, Adjusted to Actual PZR Pressure, APCF Heatup Cooldown Isothermal 50P F/hr 70°Flhr Isothermal 20°FJhr 30c Flhr 40c F/hr 50 c Flhr 100°F/hr Temperature Temperature (OF) PiJll P", Poll Pall Pall Pari P.II P.II POll (OF) (Dsla) (Dsla) (psla) (psia) (psia) (psia) (psis) (psia) (psia) 80 597 597 570 80 597 517 477 438 399 211
~o 603 601 570 90 603 525 4B6 447 409 227 100 611 601 570 611 100 534 496 458 421 246 110 621 601 570 110 621 545 508 472 436 270 120 632 601 570 120 632 559 523 469 455 299 130 646 608 570 130 646 576 542 509 477 334 140 663 622 572 150 684 644 581 140 663 597 565 534 504 377 160 710 674 597 150 684 622 592 564 537 430 I 170 741 714 621 160 710 652 626 601 577 494 180 779 7~ 654 170 741 690 667 646 628 572 190 826 626 696 160 779 736 718 101 688 664 199.9 662 682 700 190 826 192 179 168 161 157 200 869 869 738 1199.9 882 860 853 850 850 850 210 939 93~ 806 200 869 847 841 838 838 838 220 1.024 1.024 891 210 939 931 931 931 931 931 230 1.127 1.117 996 220 1.024 1,024 1.024 1.024 1,024 1,024 240 I 25<1 1211 '1 126 230 1 127 1 127 1127 1127 1 127 1127 250 1,408 1,308 1,284 240 1.254 1.254 1.254 1.254 1.254 1.254 260 1597 1429 1403 250 1,408 1,408 1408 1408 1408 1,408 270 1828 1.575 1525 280 2, 110 1,751 1674 260 1.597 1,597 1.597 1.597 1.597 1.597 290 2,454 1,971 1.856 270 1,828 1,828 1.828 1.828 1828 1,828 300 2,874 2,237 2.077 260 2.110 2.110 2.110 2.110 2,110 2.110 310 3.387 2557 2.347 290 2,454 2,454 2.454 2.454 2,454 2,454 320 2.958 2.676 300 2,874 2,874 2,874 2,874 2,874 2,874 330 3.442 3.078 310 3.367 3.387 3,387 3,367 3,387 3,223 WCAP-17l97-NP February 2010 Revision 0
2-14 Table 2-6 St. Lucie Unit 1 Cooldown and Heatup Allowable Pressure 54 EFPY, Adjusted to Indicated PZR Pressure, IPCF Heatup Cooldown Isothermal 50°Flhr 70°F/hr Isotlvlrmal 20 F/hr 0 300 FJhr .olOoFlhr 50 0 F/hr 100 0 F/hr Temperature Temperature (OF) P", P,II P,JI P,II P,n p.1I p. 1I P,II p.1I (OF) (pslal (pslal (pslal (psia) (psial (psia) lpsial (psia) (pslal 87 576 578 550 87 578 498 458 419 380 192 97 sa.! 582 550 97 584 505 467 428 390 208 107 592 582 550 107 59:1 515 477 439 40:1 227 117 601' 582 550 117 601 526 48S 453 417 251 127 613 582 550 127 613 540 504 469 435 280 137 627 588 550 147 644 603 553 137 627 557 523 490 457 315 157 665 625 562 147 644 577 545 515 465 358 167 690 655 578 157 665 602 573 545 518 411 177 722 695 602 167 690 633 607 581 558 475 187 760 746 635 177 722 6]1, 648 627 608 553 197 6{)6 606 677 167 760 717 698 682 669 645 206.9 863 863 731 197 606 772 75S 749 742 738 207 850 650 719 206.9 863 840 834 831 831 831 217 920 920 767 207 850 828 822 819 819 819 227 1.005 1.005 872 217 920 911 911 911 911 911 237 1.108 1.096 977 227 1.005 1005 1.005 1005 1.005 1005 247 1,235 1.192 1,106 237 1.106 1.100 1,108 1.108 1.100 11.108 257 1.369 1.269 1.265 247 1,235 1,235 1.235 1,235 1,235 1,235 267 1.562 1.410 1.384 257 1 369 1 369 1,369 1369 1,389 1389 277 1,793 1.540 1,506 287 2.075 1,716 1639 267 1,562 1,562 1,552 1,552 1,562 1,562 297 2,419 1,936 1,821 277 1,793 1 793 1,793 1793 1,793 1793 307 2839 2.202 2042 287 2.075 2075 2075 2075 2.075 2075 317 3352 2522 2312 297 2,419 2,419 2.419 2,419 2.419 2,419 327 2.923 2,641 307 2,839 2,636 2,836 2,836 2.836 2,839 337 3043 317 3,352 3,352 3.352 3,352 3.352 3,188 2.5 HEATUP LIMIT ANALYSIS During heatup, the thermal bending stress was compressive at the reactor vessel inside wall and was tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall and outside walls. Consequently, the outside wall, when compared to the inside wall, has the larger total stress. However, neutron embrittlement, shift in material RT NDT, and reduction in fracture toughness were greater at the inside wall. Therefore, results from both the inside and outside flaw locations must be compared to ensure the recognition of the most limiting condition. As described in the cooldown case, the reference stress intensity was calculated at the metal temperature, and the adjusted RTNUT was calculated at the tip of the flaw. Using a finite element method for the heat transfer analysis, the temperature profile through the wall and the metal temperatures at the tip of the flaw were calculated for the transient history. This information, in conjunction with the calculated wall gradient and thermal influence coefficients, was used to calculate the thermal stress intensity factor at 1/4t and 3/4t. The allowable pressure stress intensity was then determined by superpositioning the thermal stress intensity factor-with the available reference stress intensity-at the flaw tip. The allowable pressure was derived from the calculated allowable pressure stress intensity factor. A sign change occurs in the thermal stress through the reactor vessel beltline wall. Assuming there is a reference flaw at the 1/4t location, the thermal stress tends to alleviate the pressure stress, indicating that the isothermal steady-state condition represents the limiting P-T limit. However, the isothermal condition WCAP-17l97-NP February 2010 Revision 0
2-15 may not always provide the limiting P-T limit for the 1I4t location during a heatup transient. This is due to the difference between the base metal temperature and the RCS fluid temperature at the inside wall. For a given heatup rate (non-isothermal), the differential temperature through the clad and film increases as a function of thermal rate, resulting in a crack tip temperature that was lower than the RCS fluid temperature. Therefore, to ensure an accurate representation of the 1/4t P-T limit during heatup, both the isothermal and heatup rate dependent P-T limits were calculated. This also ensured that the limiting condition was recognized. These limits, in conjunction with the cooling limits, account for clad and film differential temperatures, as well as the gradual buildup of wall differential temperatures with time. To develop minimum P-T limits for the heatup transient, the isothermal conditions at 1I4t and 3/4t, 1I4t heatup, and 3/4t heatup P-T limits were compared for a given thermal rate. Then, the most restrictive P-T limits were combined, resulting in a minimum limit curve for the reactor vessel beltline for the heatup event. Table 2-7 presents the P-T results for conditions at the beltline, without applied correction factors, for isothermal, 50°FIhr, and 70°F Ihr heatup P-T limits. Table 2-5 provides the results with APCF. Table 2-6 provides results with the IPCF, which includes temperature and PCFs. Tables 2-5 and 2-6 supply the allowable PZR pressure values versus reactor coolant temperature. APCF data were compared to the design basis LTOP transient results, which are also referenced to the PZR pressure location. IPCF data were used for the recommended Technical Specification P-T limit figure changes. Uncorrected values are provided for completeness. Table 2-7 Heatup Allowable Pressures, Uncorrected Isothermal 50°F/hr 70°F/hr Temperature (OF) p.n p.n P. II (psia) (psia) (psi a) 80 657 657 629 90 663 661 629 100 671 661 629 110 680 661 629 120 692 661 629 130 706 667 629 140 723 682 632 150 744 704 641 160 769 734 657 170 801 774 681 180 839 825 714 190 885 885 756 199.9 942 942 810 200 942 942 811 210 1,012 1.012 879 220 1,097 1.097 964 230 1.200 1,190 1,069 240 1,327 1.284 1,198 250 1,481 1,381 1,357 260 1,670 1,502 1,476 270 1.901 1.648 1.598 280 2,'182 1,824 1,747 290 2,526 2,044 1,928 300 2,947 2,310 2,150 310 3,460 2,630 2,420 320 3.031 2,749 WCAP-17197-NP February 2010 Revision 0
2-16 2.6 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS Hydrostatic test limits have been calculated for 54 EFPY using the methodology of the ASME Boiler and Pressure Vessel Code, Section XI, Appendix G. The governing equation for determining the hydrostatic test limits is shown in Equation 4 from Section 2.2. The procedure was similar to calculating normal operations heatup and cooldown limits. The one exception was the factor of safety that was applied to the allowable pressure stress intensity (KIM). To account for this exception, the analysis method utilized for this calculation modified the applied factor of safety from 2.0 (for normal operation) to 1.5, for hydrostatic limits. The hydrostatic test limit establishes the minimum temperature required at the corresponding hydrostatic test pressure. Westinghouse recommends that the in-service hydrostatic test for Combustion Engineering (CE) nuclear steam supply system (NSSS) designs be performed at a test pressure corresponding to 1.1 times the operating pressure, with the reactor core not critical. Under these conditions, 10 CFR 50, Appendix G requires that the minimum temperature for the reactor vessel be at least as high as the RTNDT for the limiting material in the closure flange region, plus 90°F. However, the beltline hydrostatic test, at the recommended test pressure, has greater limitations. Therefore, it is only necessary to control plant operations to the beltline in-service hydrostatic test limits in the vicinity of this pressure. To define minimum temperature criteria for core critical operation, Appendix G of 10 CFR 50 specifies the following P-T limits. x If the RCS pressure is less than or equal to 20% of the pre-service hydrostatic test pressure (PHTP), the minimum temperature requirement for the reactor vessel must be at least as high as the RTNDT for the limiting material in the closure flange region stressed by bolt preload, plus 40°F, or the minimum permissible temperature for the in-service hydrostatic pressure test, whichever is larger. x If the RCS pressure is greater than 20% of the PHTP, the minimum temperature requirement for the reactor vessel must be at least as high as the RTNDT for the limiting material in the closure flange region stresses by bolt preload, plus 160°F, or the minimum permissible temperature for the in-service hydrostatic pressure test, whichever is larger. According to Appendix G to 10 CFR 50, the following calculation specifies P-T limits for core critical operation to provide additional margin during actual power operation. In-service hydrostatic pressure =
= (1.1 x operating pressure) + instrumentation uncertainty
= (1.1 x (2,250-15) + 15 psia) + 0 psi = 2,473.5 psia Pressure instrumentation uncertainty was not included. Furthermore, the factor 1.1 was used for the gauge units (psig) of operating pressure instead of the absolute units (psia).
WCAP-17197-NP February 2010 Revision 0
2-17 The minimum temperature for the core critical operation and the hydrostatic test was the temperature corresponding to the in-service hydrostatic pressure. The minimum temperature for the hydrostatic and leak test cases was 270.7°F. This temperature value was obtained from Table 2-8 (unadjusted, beltline data) by interpolating the temperature values to the pressure given in the equation above. Hydrostatic test limits are tabulated in Table 2-8, and are adjusted using the correction factors for both the APCF and IPCF cases. For both the APCF and IPCF cases, the specified beltline heatup P-T limit was more restrictive at temperatures above 270.7°F and 277.7°F respectively. Consequently, the core critical limits have been established as a combination of this temperature and the specified heatup P-T limit from ASME Appendix G, plus 40°F. The core critical limits established were based solely on fracture mechanics considerations and do not consider core physics safety analyses. Core physics safety analyses can control the temperature at which the core can be brought critical. Table 2-8 St. Lucie Unit 1 Hydrostatic Test P-T Limit Data Actual Pressurizer, Indicated Pressurizer, Beltline (Uncorrected) APCF Conditions IPCF Conditions Conditions Tellllpemtur,e P 1 empelmture, P Temp lila:! (OF) (IPSJild (OF) (psJilll' (1IJiSIia)
'I'" F) 80 811 81 79'2 80 811 90 820 91 800 90 819' 100 830 101 81"1 100 890 110 842 111 823 '110 902 120 858 121 83S 12D 917 130 876 137 857 no 936 140 800 147 880 14'0 959 150 927 157 908 '150 987 160 961 161 942 160 1 021 110 1,003 171 9184 170 1 063 180 1.054 181 1034 180 1.113 190 1.116 197 1,091 190 1.176
'199.9 1.191 206.9 1.172 199.9 1 251 200 1,119 201 1,159 200 1251 210 1..211 211 1,,252 210 1 344 220 1,,384 227 1,365 220 1457 230 1,522 231 1,503 230 1 595 240 1,,691 241 1,'656 24J0 1 764 250 1.891 251 U\62 250 1 970 260 2.149 267 2,114 260 2222 268.2 2,41!J0.7 275.2 2.,365.7 270 2,529 550 2,400.7 550 2,365.7 280 2.905 290 3363 WCAP-17197-NP February 2010 Revision 0
2-18 2.7 LOWEST SERVICE TEMPERATURE, MINIMUM BOLTUP TEMPERATURE, FLANGE LIMIT TEMPERATURE, MINIMUM PRESSURE LIMITS AND LTOP ENABLE TEMPERATURES In addition to the computation of the reactor vessel beltline P-T limits, additional limits have been provided for reference. These additional limits were the lowest service temperature (LST), minimum bolt-up temperature, flange limit temperature, and minimum pressure limit. LST is defined in ASME Section III, NB-3211 as the minimum temperature for piping, pumps, and valves (the remainder of the RCS) in the RCS in order to exceed the 20% of the pre-service hydrostatic test pressure. The LST is established as a temperature not less than RTNDT of the remainder of the RCS plus 100°F. Previously, an RTNDT of 90°F had been applied in such calculations for St. Lucie Unit 1. It was found that this limitation was associated with an estimate related to the RCP materials. It was determined that the RCP pump shaft, casing, casing wear ring, hydrostatic bearing, and pump cover are made of stainless steel and, therefore, do not affect the limiting RTNDT. The next most limiting RTNDT documented for the RCS piping was 58°F. Therefore, the LST was 158°F. When the pressure exceeds 20% of pre-service hydrostatic test pressure, the temperature of the closure flange regions must exceed the initial RTNDT of the material by at least 120°F for normal operation, and by 90°F for hydrostatic and leak testing. The minimum pressure limit is applicable between the minimum bolt-up temperature, lowest service temperature, and the flange limit temperature. Defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure, the minimum pressure is as follows. 20% of pre-service hydrostatic test = (1.25 x design pressure) x 0.20
= 1.25 x (2,500-15) x 0.20 + 15 = 636.25 psia With the correction factors as developed in Section 2.2.1 (PAPCF = 59.8 psid, PIPCF = 79.0 psid), this pressure was adjusted to 576.5 psia for APCF, and 557.3 psia for the IPCF cases.
The scale factor used on the design pressure in the previous calculation was the gauge value (psig) instead of the absolute pressure (psia). The minimum bolt-up temperature was defined as 80°F, which provides margin to protect the vessel head, vessel flange, and upper shell from being stressed at a temperature below the RTNDT of those materials. The P-T curves include a 7°F margin shift for indicated instrument uncertainty so that the operator does not need to account for the instrument error at bolt-up. For steady state, a 30°F margin on minimum bolt-up temperature was already in place since the lowest RTNDT of the flange region was 50°F. Low Temperature Overpressure Protection (LTOP) enable temperatures are determined per ASME Boiler and Pressure Vessel Code Section XI, Appendix G. The Code states that the LTOP systems become effective at coolant temperatures less than 200°F, or at coolant temperatures corresponding to RV temperatures less than RTNDT + 50°F, whichever is greater. The LTOP enable temperature for cool-down is based on the isothermal pressure-temperature (P-T) limit. For cool-down, including instrumentation uncertainty (IPCF case assumed): WCAP-17197-NP February 2010 Revision 0
2-19 LTOP enable temperature= RTNDT + 50°F
= 210°F + 50°F + 7°F = 267°F For heat-up transients with a 70°F/hr rate, the coolant temperature that corresponds to the crack tip temperature of RTNDT + 50°F = 260°F (from the heat transfer analysis results) is 291.9°F. With instrument uncertainty added, it is 298.9°F (IPCF assumed). Details of LTOP enable temperatures are given in Table 2-9.
Table 2-9 LTOP Enable Temperature Limits Uncorrected LTOP Case Tcoolant Enable (°F) (°F) HU 10°F/hr 264.6 271.6 HU 20°F/hr 269.2 276.2 HU 30°F/hr 273.8 280.8 HU 40°F/hr 278.4 285.4 HU 50°F/hr 283.0 290.0 HU 60°F/hr 287.5 294.5 HU 70°F/hr 291.9 298.9 CD / Isothermal 260.0 267.0 WCAP-17197-NP February 2010 Revision 0
2-20 2.8 DATA Reactor Vessel Data Reference Design Pressure = 2500 psia 16 Design Temperature = 650°F 16 Operating Pressure = 2250 psia 16 Beltline Thickness = 8.625 in 16 Inside Radius = 86.914 in 16 Outside Radius = 95.85 in 16 Cladding Thickness = 0.3125 in 16 Material SA-533-65 Grade B Reference Thermal Conductivity = 23.8 BTU/hr-ft-°F 13 Youngs Modulus = 28 x 106 psi 13 Coefficient of Therma1 Expansion = 7.8 x 10-6 in/in-°F 13 Specific Heat = 0.122 BTU/lb-°F 13 Density = 490 lb/ft3 13 Material SA-533-65 Grade B Reference Thermal Conductivity = 23.8 BTU/hr-ft-°F 13 Youngs Modulus = 28 x 106 psi 13 Coefficient of Therma1 Expansion = 7.8 x 10-6 in/in-°F 13 Specific Heat = 0.122 BTU/lb-°F 13 Density = 490 lb/ft3 13 Stainless Steel Cladding Reference Thermal Conductivity = 10.1 BTU/hr-ft-°F 13 Film coefficient on inside surface = 1000 BTU/hr-ft2-°F Assumption WCAP-17197-NP February 2010 Revision 0
2-21 Pressure Correction Factors for Elevation and Flow as Developed in Section 2.2.1 Applicable to all plant condition with two or less RCP in operation: APCF APCF = 59.8 psid Indicated pressure correction factor: Narrow-range pressure instruments: IPCF = 79.0 psid Wide-range pressure instruments: IPCF = 94.8 psid Corresponding information values for three or less operating RCP: APCF APCF = 72.8 psid Indicated pressure correction factor: Narrow-range pressure instruments: IPCF = 92.0 psid Wide-range pressure instruments: IPCF = 107.8 psid WCAP-17197-NP February 2010 Revision 0
3-1 3 LOW-TEMPERATURE OVERPRESSURE PROTECTION 3.1 GENERAL The primary objective of the LTOP system is to preclude the violation of applicable Technical Specification P-T limits during startup and shutdown conditions. These P-T limits were usually applicable to a finite time period of operation and were based upon the irradiation damage prediction by the end of the period. Accordingly, each time new P-T limits become effective, the LTOP system needs to be re-analyzed and modified, if necessary, to continue its function. The LTOP system prevents the violation of the RCS brittle fracture P-T limits in the event of an overpressure event within the LTOP temperature range. An RCP start overpressure event is one of two design basis events for the LTOP system. The RCP start is referred to as the energy addition event. The other design basis event is the mass addition transient, which is typically based on an inadvertent safety injection actuation signal (SIAS) in the LTOP temperature range. A typical LTOP system includes pressure-relieving devices and a number of administrative and operational controls. At St. Lucie Unit 1, the current LTOP system uses two power-operated relief valves (PORVs) for the LTOP temperature range from the minimum bolt-up temperature, to the LTOP enable temperature. The PORVs (tag numbers V1402 and V1404) have two opening setpoints of 350 and 530 psia. These relief valves, in combination with certain other limiting conditions for operation contained in Technical Specifications, comprise the St. Lucie Unit 1 LTOP system. Since the new P-T limits described in this report cover the operating period ending at 54 EFPY, the existing LTOP system was re-analyzed to determine if modifications are required or improvements can be implemented in order for the system to provide adequate LTOP through 54 EFPY. The LTOP system was analyzed for the expected conditions following implementation of the EPU. 3.2 METHOD AND ASSUMPTIONS The approach taken in performing the LTOP evaluation was to analyze the existing PORV setpoints. Accordingly, the existing PORV setpoints of 350 psia and 530 psia were used. The following existing general assumptions were used in the LTOP analyses.
- 1. Only one PORV is available.
- 2. The RCS is in a water solid condition.
- 3. The letdown flow paths are isolated.
- 4. The PZR heater input and decay heat input was considered as additional energy sources.
- 5. There is no heat absorption or metal expansion at the primary pressure boundaries.
WCAP-17197-NP February 2010 Revision 0
3-2 The PORV opening characteristics were adjusted for control circuit uncertainty and valve response time. This was addressed in the following manner.
- 1. The RCS pressure just prior to PORV opening was conservatively assumed to be greater than the nominal PORV setpoint, because of the relative pressure instrument uncertainty between the pressure indication and the PORV actuation channels. This 26 psi uncertainty is provided in Reference 14.
- 2. The PORV opening time was previously assumed to equal 0.25 seconds, which enveloped the opening times observed in applicable tests. Based on an evaluation of the test data, it was conservatively assumed that this total opening time was a better indication of the PORV stroke time. The computer code that was used to model the energy addition transient could not model a ramped PORV opening. To account for a ramp opening during stroke time, a delay in the PORV opening equal to a sum of a conservative solenoid delay time of 0.65 seconds, and one half of the previously discussed stroke time (0.125 seconds), were assumed in the energy addition transient analysis. This delay was assumed to be followed by instantaneous opening. The PORV opening setpoint used in the energy addition transient analysis code was adjusted based on this delay time, assuming a bounding pressure ramp rate prior to valve opening. In the mass addition transient analysis, the PORV was opened in time steps following the solenoid delay. The product of the PORV capacity and the time passed over the stroke time was credited as the stroke time passes, until the PORV was full open.
- 3. The impact of the PORV opening time was taken into account in the energy addition transient analysis by adding transient-specific pressure accumulation during 0.775 seconds (0.65 seconds plus 0.125 seconds) to the opening pressure to arrive at the maximum opening pressure. Pressure accumulation was assumed to be a function of an applicable pressurization ramp rate moments prior to reaching the valve setpoint.
Based on the existing analyses, modified assumptions and inputs, and new maximum transient pressures for the same design basis transients were determined as appropriate. Out of these, the most limiting pressures in given temperature ranges were selected as controlling the limiting temperatures for LTOP. Finally, by comparing these controlling pressures to the P-T limit curves for 54 EFPY, limiting conditions for operation were identified. 3.3 PRESSURE TRANSIENT ANALYSES 3.3.1 Energy Addition Transients The energy addition analysis determined the peak pressure that would occur as a result of the RCS pressure transient caused by an RCP start with an initial steam generator-to-reactor-vessel temperature differential of 30°F (Reference 15, LCO 3.4.1.4.1), during RCS water-solid, low-temperature conditions. PORVs, in accordance with Reference 15, LCO 3.4.9.1 and LCO 3.4.13, provide LTOP system overpressure protection at St. Lucie Unit 1. This calculation analyzed cases with a single PORV providing LTOP overpressure protection. WCAP-17197-NP February 2010 Revision 0
3-3 An analysis methodology, consistent with the transient analysis of record, was used. Plant-specific volumes, masses, decay heat, RCP heat, PZR heater contributions, selected initial temperatures, and heat transfer coefficients were incorporated into the analysis input, which produced results in the form of RCS system pressure values versus time. The PORV mitigated energy addition transient was analyzed for the existing PORV setpoints of 350 psia and 530 psia (Reference 15, LCO 3.4.13). The analysis assumed that the pressure transient was taking place in the PZR. The effect of the PORV inlet piping on the analysis results was taken into account by determining PORV flow rates at PORV inlet pressure, which was corrected from PZR pressure for elevation difference and flow losses. This correction reduces PORV discharge, thus maximizing the transient pressures. The following major assumptions were used in the analysis of the RCP start transient, in addition to the assumptions mentioned above and in Section 3.2.
- 1. The PORV opening occurs at an opening pressure that is greater than the nominal setpoint by a sum of the pressure instrument uncertainty and pressure accumulation due to finite opening time.
This assumption maximizes RCS pressure at the PORV opening.
- 2. The cases were analyzed for initial RCS fluid temperatures of 140°F, 200°F, and 300°F. These temperatures were consistent with those assumed in existing analyses, as well as with the updated LTOP enable temperature 300°F for cooldown (Table 3-1).
- 3. The initial RCS pressure is 300 psia, which is consistent with existing analyses.
- 4. The historical St. Lucie Unit 1 LTOP energy addition analyses only consider the water and metal masses in the region of the tube bundle to contribute as heat sources. This analysis maintains the assumption.
- 5. The RCP heat input is considered as an additional energy source.
The PORV mitigated pressure transient at the 350 psia setpoint is illustrated in Figures 3-1 and 3-2. The resulting maximum transient pressures (adjusted to PZR pressures) of 420 psia and 393 psia at 200°F and 140°F, respectively, are provided in Table 3-2. The PORV mitigated pressure transients at the 530 psia setpoint is illustrated in Figures 3-3 and 3-4 and the resulting maximum transient pressure (adjusted to PZR pressure) of 580 psia is provided in Table 3-1. WCAP-17197-NP February 2010 Revision 0
3-4 Table 3-1 Maximum Transient Pressures at 530 psia Setpoint Transient Applicable Transient PORV Setpoint 530 psia Type HU or CD T (°F) P (psia) 2 HPSI 3 CP 300 1080 2 HPSI 3 CP 220 1048* 1 HPSI 3 CP 300 834 Mass 1 HPSI 3 CP 220 723* Addition 3 CP 220 570 Transients 3 CP 140 570 1 HPSI 220 595 1 HPSI 140 591 Energy RCP Start 300 580 Addition RCP Start 200 580 Transients
- Transient pressures are provided for temperatures that envelope the range of applicability. TS LCO 3.5.3 (Reference 15) requires that a maximum of one HPSI pump be operable below 270°F and that all HPSI pumps be disabled below 236°F unless, as specified by TS LCO 3.1.2.1 and 3.1.2.3 (Reference 15), one HPSI pump is established to ensure boration capability and all CP are disabled.
Table 3-2 Maximum Transient Pressures at 350 psia Setpoint Transient Applicable Transient PORV Setpoint 350 psia Type HU or CD T (°F) P (psia) 2 HPSI 3 CP N/A N/A 2 HPSI 3 CP N/A N/A 1 HPSI 3 CP N/A N/A Mass 1 HPSI 3 CP N/A N/A Addition 3 CP 200 392 Transients 3 CP 140 392 1 HPSI 200 595 1 HPSI 140 500 Energy RCP Start 200 420 Addition RCP Start 140 393 Transients WCAP-17197-NP February 2010 Revision 0
3-5 Case 1 RCS Pressure vs. Time 405 390 375 RCS Pressure, PSIA 360 345 330 315 300 0 5 10 15 20 25 Time, Sec Figure 3-1 St. Lucie 1, Energy Addition Transient Case 1, PORV, PSET = 350 psia, TC = 140°F Case 2 RCS Pressure vs. Time 435 420 405 RCS Pressure, PSIA 390 375 360 345 330 315 300 0 5 10 15 20 25 Time, Sec Figure 3-2 St. Lucie 1, Energy Addition Transient Case 2, PORV, PSET = 350 psia, TC = 200°F WCAP-17197-NP February 2010 Revision 0
3-6 Case 3 RCS Pressure vs. Time 600 550 RCS Pressure, PSIA 500 450 400 350 300 0 5 10 15 20 25 Time, Sec Figure 3-3 St. Lucie 1, Energy Addition Transient Case 3, PORV, PSET = 530 psia, TC = 200°F Case 4 RCS Pressure vs. Time 600 550 RCS Pressure, PSIA 500 450 400 350 300 0 5 10 15 20 25 Time, Sec Figure 3-4 St. Lucie 1, Energy Addition Transient Case 4, PORV, PSET = 530 psia, TC = 300°F WCAP-17197-NP February 2010 Revision 0
3-7 3.3.2 Mass Addition Transients The RCS pressure transient due to an inadvertent safety injection actuation was the design basis mass addition transient. The most severe mass addition transient occurs due to simultaneous actuation of two high-pressure safety injection (HPSI) pumps and three charging pumps (CPs) while letdown is isolated. This transient, however, was only analyzed at RCS temperature above 270°F, consistent with existing LTOP controls on HPSI pump availability limitations in the Reference 15, LCO 3.5.3. As a result, at RCS temperature below 270°F, the most limiting mass addition transient was due to one HPSI and three CPs input. The following major assumptions were used in the analysis of the mass addition transients, in addition to the assumptions mentioned above and in Section 3.2.
- 1. It was assumed that the shut down cooling system (SDCS) will be aligned below 200qF. In this configuration, one HPSI and three CPs may be aligned.
- 2. The configuration with the SDCS isolated may allow two HPSI pumps and three CPs to be aligned. The PORV is the primary LTOP protection device.
- 3. In all transient cases, only a single pressure protection relief valve was assumed.
- 4. As many as three RCPs were operational at startup and during fill and vent and could be operating during the LTOP mass addition transient. However, the RCP heat input for the mass addition transient (consistent with current methodology) need not be considered since the transient initiates with the plant in a steady-state condition (operator controlled heatup or cooldown) and instantaneous RCP start is not a credible transient input.
- 5. PZR initial conditions were assumed to be 500 psia, 260 psia, and 75 psia for RCS hydraulic temperatures of 300°F, 220°F and 140°F respectively. The PZR was assumed to be saturated in each condition with temperatures of 467°F, 404°F and 308°F respectively, consistent with existing analyses.
The analysis updates the existing design inputs and assumptions to more accurately represent the current operating configuration. RCS volume expansion due to contributions from decay heat and full PZR heater heat were taken into account. PORV discharge flowrates as a function of PZR pressure are plotted in Figures 3-5 through 3-12. The mass addition events (including the RCS volume expansions) are compared to the PORV (Figures 3-5 through 3-12) cases and equilibrium pressures were determined. An equilibrium pressure is the pressure at which the mass inputs match the relief valve discharge. PORV transient analyses were performed to determine maximum transient pressures for both the PORV set pressures of 350 psia and 530 psia. The transient analysis calculated RCS pressure over time steps until an equilibrium was reached between HPSI and CPs, inflow and PORV outflow. WCAP-17197-NP February 2010 Revision 0
3-8 The equilibrium pressures relevant at a PORV setpoint of 530 psia were as follows. Equilibrium Pressure (psia) Transient (PORV Mitigation at 530 psia Setpoint) 300°F 220°F 140°F 2 HPSI + 3 CPs 1080 1048 1 HPSI + 3 CPs 834 723 3 CPs 286 113 Single HPSI 595 521 The equilibrium pressure was limiting in most of these cases. However, similar to in the case of the energy transient, pressure accumulation prior to the opening of the PORV can exceed the equilibrium pressure. This occurs during the transient specific to the three CPs, as well as for the lower temperature range of the single HPSI transient. The equilibrium pressure for the single HPSI case bounds the peak opening pressure for the three CP case and therefore the equilibrium pressure remains limiting. The maximum opening pressure for the single HPSI case was 591 psia, and therefore this value was used in place of the 140°F value of 521 psia. The equilibrium pressures relevant at a PORV setpoint of 350 psia were as follows: Equilibrium Pressure (psia) Transient (PORV Mitigation at 350 psia Setpoint) 220°F 140°F 3 CPs 286 113 Single HPSI 595 521 The equilibrium pressure associated with the single HPSI transient was limiting for all cases. The final results of the mass addition transient analysis are provided in Tables 3-1 and 3-2. WCAP-17197-NP February 2010 Revision 0
3-9 Case 1 1085 1084 1083 1082 PZR Pressure (psia) PORV Discharge 1081 Flowrate 1080 1079 1078 1077 Mass Input 1076 1075 935 937 939 941 943 945 Flow Rate (gpm) Figure 3-5 St. Lucie Unit 1 LTOP Mass Addition Transient Case 1 2 HPSI + 3 CPs, 300°F WCAP-17197-NP February 2010 Revision 0
3-10 Case 3 1050 1049 1048 1047 PZR Pressure (psia) 1046 1045 Mass Input 1044 1043 1042 PORV Discharge 1041 Flowrate 1040 930 935 940 945 950 Flow Rate (gpm) Figure 3-6 St. Lucie Unit 1 LTOP Mass Addition Transient Case 3 2 HPSI + 3 CPs, 220°F WCAP-17197-NP February 2010 Revision 0
3-11 Case 4 840 839 838 837 PZR Pressure (psia) PORV Discharge 836 Flowrate 835 834 833 832 Mass Input 831 830 740 745 750 755 760 Flow Rate (gpm) Figure 3-7 St. Lucie Unit 1 LTOP Mass Addition Transient Case 4 1 HPSI + 3 CPs, 300°F WCAP-17197-NP February 2010 Revision 0
3-12 Case 6 725 724 723 722 PZR Pressure (psia) 721 720 719 PORV Discharge 718 Flowrate 717 Mass Input 716 715 750 752 754 756 758 760 Flow Rate (gpm) Figure 3-8 St. Lucie Unit 1 LTOP Mass Addition Transient Case 6 1 HPSI + 3 CPs, 220°F WCAP-17197-NP February 2010 Revision 0
3-13 Case 7 290 289 288 287 PZR Pressure (psia) 286 285 284 PORV Discharge 283 Flowrate Mass Input 282 281 280 250 252 254 256 258 260 Flow Rate (gpm) Figure 3-9 St. Lucie Unit 1 LTOP Mass Addition Transient Case 7 3 CPs, 220°F WCAP-17197-NP February 2010 Revision 0
3-14 Case 8 120 119 118 117 PZR Pressure (psia) 116 PORV Discharge 115 Flowrate 114 113 112 Mass Input 111 110 210 215 220 225 230 Flow Rate (gpm) Figure 3-10 St. Lucie Unit 1 LTOP Mass Addition Transient Case 8 3 CPs, 140°F WCAP-17197-NP February 2010 Revision 0
3-15 Case 9 600 599 598 597 PZR Pressure (psia) PORV Discharge 596 Flowrate 595 594 593 592 Mass Input 591 590 650 655 660 665 670 Flow Rate (gpm) Figure 3-11 St. Lucie Unit 1 LTOP Mass Addition Transient Case 9 Single HPSI Pump, 220°F WCAP-17197-NP February 2010 Revision 0
3-16 Case 10 525 524 523 522 PZR Pressure (psia) PORV Discharge 521 Flowrate 520 519 518 517 Mass Input 516 515 640 645 650 655 660 Flow Rate (gpm) Figure 3-12 St. Lucie Unit 1 LTOP Mass Addition Transient Case 10 Single HPSI Pump, 140°F WCAP-17197-NP February 2010 Revision 0
3-17 3.3.3 Controlling Pressures The pressure transient analysis results contained in Tables 3-1 and 3-2 were evaluated to identify the controlling pressures and applicable temperature ranges. The controlling pressures were the maximum transient pressures of all applicable transients in a particular temperature region. The maximum pressure was determined for each transient by conservative interpolation for the temperature range pertinent to the specific transient. The maximum pressures for the range of temperatures were used to determine the appropriate limiting conditions for operation. These limiting conditions for operation are provided in Section 3.4. 3.4 LIMITING CONDITIONS FOR OPERATION The temperature requirements for selecting the setpoints for the PORVs for LTOP and the limitations on heatup and cooldown rates are provided in Table 3-3. These requirements were based on PORV setpoints of 350 and 530 psia. An LTOP enable temperature of 300°F for both heatup and cooldown is conservative with respect to the values presented in Table 2-9. This conservative approach, especially with respect to the cooldown limit of 267°F, is appropriate from the human performance perspective as it provides operational consistency and simplicity. It should be noted that during heatup, the PORV setpoint can be changed to 530 psia at any temperature above the minimum cold leg PORV setpoint transition temperature of 200°F in Table 3-3. During cooldown the PORV setpoint must be changed to 350 psia before or upon reaching the indicated temperature of 200°F in Table 3-3. The existing Technical Specification LTOP requirements related to the limitations on RCP starts, operating RCP and HPSI pump alignment to the RCS remain unchanged except for the temperature range of applicability for the RCP start limitations as well as the elimination of the HPSI throttling requirements and case specific heatup and cooldown rates for HPSI alignment below the 236°F limit due to failure of all three CPs. Table 3-3 LTOP Requirements, 54 EFPY Low-temperature RCS Overpressure Protection Range Operating Cold Leg Temperature, °F Period, EFPY During Heatup During Cooldown
< 54 < 300 < 300 Minimum Cold Leg Temperature for PORV Setpoint Transition for LTOP Operating Cold Leg Temperature, °F Period, EFPY During Heatup During Cooldown
< 54 200 200 WCAP-17197-NP February 2010 Revision 0
3-18 Table 3-3 LTOP Requirements, 54 EFPY (cont.) Maximum Allowable Heatup Rates 70°F /hr, at all temperatures Maximum Allowable Cooldown Rates 20°F /hr, at Tc < 125°F 30°F /hr, at Tc > 125°F 40°F /hr, at Tc > 145°F 50°F /hr, at Tc > 160°F 100°F /hr, at Tc > 180°F Note: The applicability of the following restrictions is established as TC < 300°F (This is a modified applicability band.) x A RCP shall not be started with two idle loops, unless the secondary water temperature of each steam generator is less than 30°F above each of the RCS cold leg temperatures. (This is an existing limitation.) x Prior to decreasing the RCS temperature below 270°F, a maximum of only one HPSI pump shall be operable with its associated header stop valve open. (This is an existing limitation.) x Prior to decreasing the RCS temperature below 236°F, all HPSI pumps shall be disabled and their associated header stop valves closed except in the case where all CPs have failed. In this case, the previous single HPSI limitation remains, with the added restriction that all CPs shall be disabled. (This is an existing limitation.) 3.5
SUMMARY
OF PROPOSED CHANGES The proposed LTOP system is designed in accordance with the requirements set forth in the NRC Branch Technical Position BTP 5-2, contained within SRP 5.2.2, Reference 5. The proposed system is adequate to prevent violation of Appendix G P-T limits during the operating period ending at 54 EFPY. In order to implement the proposed LTOP system the following is required: x Modification of appropriate Technical Specifications x Modification of appropriate plant operating procedures The implementation of the proposed LTOP system will not result in a reduction in the margin of safety presently afforded by Technical Specifications. WCAP-17197-NP February 2010 Revision 0
4-1 4 REFERENCES
- 1. Code of Federal Regulations, 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, January 2006.
- 2. Code of Federal Regulations, 10 CFR 50, Appendix G, Fracture Toughness Requirements, December 1995.
- 3. ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Fracture Toughness Criteria for Protection against Failure, 2002 Edition with the 2003 Addenda.
- 4. Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, Revision 2, May 1988.
- 5. U. S. Nuclear Regulatory Commission Standard Review Plan (SRP) 5.2.2, Overpressure Protection, Revision 3, March 2007.
- 6. Florida Power and Light Letter, L-2004-244, St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Extension of the Reactor Coolant System Pressure/Temperature Curve Limits and LTOP to 35 EFPY, December 20, 2004.
- 7. Florida Power and Light Letter, L-97-136, St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 NRC TAC Nos. M95484 and M95485 Request for Additional Information - Response to 10 CFR 50.61 - Pressurized Thermal Shock Evaluation, May 16, 1997.
- 8. Florida Power and Light Letter, L-97-10, St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 NRC TAC Nos. M95484 and M95485 Request for Additional Information (RAI) -
Response to 10 CFR 50.61 - Pressurized Thermal Shock Evaluation, January 14, 1997.
- 9. NRC Reactor Vessel Integrity Database, Version 2.0.1 (RVID-2), July 6, 2000.
- 10. ASME Boiler and Pressure Vessel Code Section XI, Appendix A, Analysis of Flaws, 1998 Edition with the 2000 Addenda.
- 11. Westinghouse Report, CE-NPSD-683-A Task-1174, Revision 06, Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications, April 2001.
- 12. J. Heliot, R.C. Labbens, and Pellisser-Tanon, Semi-Elliptical Cracks in a Cylinder Subjected to Stress Gradients, ASTM Special Technical Publication 677, August 1979.
- 13. ASME Boiler and Pressure Vessel Code Section III, Appendix I, Design Stress Intensity Values, Allowable Stresses, Material Properties, and Fatigue Design Curves, 1989 Edition.
- 14. FPL Calculation, PSL-1FJI-09-001, Revision 0, Instrument Uncertainty Calculation for Pressure - Temperature Limit Curves and LTOP St Lucie Unit 1, April 2009.
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4-2
- 15. St. Lucie Plant Unit No. 1 Technical Specifications, Amendment 204, February 22, 2008.
- 16. CE Report Pressure-Temperature Limits and Low Temperature Overpressure Protection for St.
Lucie Unit 1 for 15 Effective Full Power Years, Revision 1, September 1989. Transmitted to FPL via Letter F-MPS-89-046, P. J. Hijeck to T. E. Roberts, September 27, 1989. WCAP-17197-NP February 2010 Revision 0
A-1 APPENDIX A TECHNICAL SPECIFICATION FIGURES WCAP-17197-NP February 2010 Revision 0
A-2 FIGURE 3.4-2a ST. LUCIE UNIT 1 PIT LIMITS, 54 EFPY HEATUP AND CORE CRITICAL 2000 (f) D... W 0:::: ISOTHERMAL (f) (f) 1500 w 0:::: D... 0:::: ~LOWEST SERVICE TEMPERATURE 165°F W N 0:::: /to.. (f) (f) w 1000 CORE CRITICAL 0:::: D... I 0 0 70 F/HR W I- << -y.... I 0 ./ 0 z j550.0 PSIA (J) u 500
~
D... ALLOWABLE HEATUP RATE ~ MIN. BOLTUP TEMP. 80°F 70°F/HR f o o 100 200 300 400 500 T c - INDICATED RCS TEMPERATURE,oF WCAP-17l97-NP February 2010 Revision 0
A-3 FIGURE 3.4-2b S1. LUCIE UNIT 1 PIT LIMITS, 54 EFPY COOLDOWN AND INSERVICE TEST INSERVICE HYDROSTATIC TEST (j) 100°F/HR TO ISOTHERMAL 0... w-0:::
=>
(j) (j) 1500 w 0::: 0... 0::: ~LOWEST SERVICE TEMPERATURE 165°F W N 0:::
=>
(j) (j) W 0::: 0... o W
~
ISOTHERMAL
<<o CJ5573 PSIA o
z en I--- 20°F/HR U 0::: 500 I--- I--- 30°F/HR 0... I--- I--- 40°F/HR ALLOWABLE COOLDOWN RATES I 50°F/HR RATE °F/HR TEMP. LIMIT of 20 <125 30 125-145 C==:J100°F/HR 40 145-160 50 160-180 MIN. BOLTUP TEMPERATURE 80°F ~ 100 >180 0 0 100 200 300 400 500 T c -INDICATED RCS TEMPERATURE,oF WCAP-17l97-NP February 2010 Revision 0}}