ML040630675
| ML040630675 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/29/2004 |
| From: | Bamford W, Kaihwa Hsu, Petsche J, Swamy S, Yang C Westinghouse |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WCAP-16142-NP, Rev 1 | |
| Download: ML040630675 (149) | |
Text
{{#Wiki_filter:Enclosure 6 Vogtle Electric Generating Plant Units I and 2 Non-proprietary version of WCAP-16142-NP, Rev. I
Westinghouse Non-Proprietary Class 3 WCAP-1 6142-NP February 2004 Revision 1 Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and 2 Westinghouse
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16142-NP Revision 1 Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and 2 Warren Bamford K. Robert Hsu Joseph F. Petsche February 2004 Reviewer: (_ _ _ __ C. Y. Yang Piping Analysis and Fracture Mechanics Approved:. Piping Analysis and Fracture Mechanics Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355 0 2004 Westinghouse Electric Company LLC All Rights Reserved 6287-NP.doc-2/1 6/04
iii TABLE OF CONTENTS I INTRODUCTION ................................................... 1-I 2 TECHNICALAPPROACH ................................................... 2-1 3 FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES . .................................. 3-1 3.1 STRESS INTENSITY FACTOR CALCULATIONS ................................................... 3-1 3.2 FRACTURE TOUGHNESS ................................................... 3-1 3.3 IRRADIATION EFFECTS ................................................... 3-2 4 FLANGE INTEGRITY .................................................... 4-1 5 ARE FLANGE REQUIREMENTS NECESSARY? ................................................... 5-1 6 SAFETY IMPLICATIONS OF THE FLANGE REQUIREMENT .............................................. 6-1 7 REFERENCES ................................................... 7-1 APPENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY*. .A-I APPENDIX B THERMAL AGING OF FERRITIC RPV STEELS AT REACTOR OPERATING TEMPERATURES . 1 l-APPENDIX C STRESS DISTRIBUTIONS IN THE CLOSURE HEAD REGION . C- I APPENDIX D FLANGE INSPECTION RESULTS: VOGTLE UNITS . 1. 1 February Revision2004I 16142-NP WCAP- I161 42-NP February 2004 6287-NP.doc-211 6AM 6287-NP.doc-2/116/04 Revision I
1-1 1 INTRODUCTION 10 CFR Part 50, Appendix G contains requirements for pressure-temperature limits for the primary system, and requirements for the metal temperature of the closure head flange and vessel flange regions. The pressure-temperature limits are to be determined using the methodology of ASME Section XI, Appendix G [1], but the flange temperature requirements are specified in IOCFR5O Appendix G This rule states that the metal temperature at the closure flange regions must exceed the material unirradiated RTNwr by at least 1200 F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure, which is 621 psig for a typical PWR, and 300 psig for a typical BWR. This requirement was originally based on concerns about the fracture margin in the closure flange region. During the boltup process, outside surface stresses in this region typically reach over 70 percent of the steady state stress, without being at steady state temperature. The margin of 120 0 F and the pressure limitation of 20 percent of hydrotest pressure were developed using the Kl,, fracture toughness, in the mid 1970s, to ensure that appropriate margins would be maintained. Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of K1 , in the development of pressure-temperature curves. as contained in ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1." ASME Code Case N-640 was approved for use without conditions by the NRC in Regulatory Guide 1.147 [16]. Figure 1-1 illustrates the problem created by the flange requirements for a typical PWR heatup curve. ft is easy to see that the heatup curve using K1 , provides for a much higher allowable pressure through the entire range of temperatures. For this plant, however, the benefit is negated at temperatures below RTII, +120'F because of the flange requirement of 10 CFR Part 50, Appendix G The flange requirement ol 10 CFR 50 was originally developed using the Kia fracture toughness. and this report will show that up oX1 the newly accepted K1, fracture toughness for flange considerations leads to the conclusion that the ll.auiv requirement can be eliminated for Vogtle Units I and 2. Revision 1. Created to correct errata in the report. No technical changes were made. WCAP-16142-NP February 2004 6287-NP.doc-2116104 Revision I
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35D Temprtu, (F) Figure 1-1 Illustration of the Impact of the Flange Requirement for a Typical PWR Plant WCAP- I6142-NP February 2004 6287-NP.doc-2/16/04 Revision I
2-1 2 TECHNICAL APPROACH The evaluation presented here is intended to cover the Vogtle Units 1 and 2 reactor vessels. Fracture evaluations have been performed on the closure head geometry specific to these units, and results will be tabulated and discussed. The geometry of the closure head region for Vogtle Units I and 2 is shown in Figure 2-1. Stress analyses have been performed, and these stress results were used to perform fracture mechanics evaluations. Details of the finite element stress analysis results are provided in Appendix C. The highest stress location in the closure head and vessel flange region is in the head, just above the bolting flange. This corresponds with the location of two welds as shown in Figure 2-1. The highest stressed location is near the outside surface of the head in that region, and so the fracture evaluations have assumed a flaw at the outside surface. The goal of the evaluation is to compare the structural integrity of the closure head during the boltup, plant heatup and plant cooldown processes, to the structural integrity during steady state operation. The question to be addressed is: With the higher Kc, fracture toughness now known to be applicable, is there still a concern about the structural integrity of the closure head during boltup? WCAP-1 6142-NP February 2004 6287-NP.doc-2126/04 Revision I
2-2 TOP HEAD DOME TO TORUS WELD
, (CUT 2) 11-S TORUS TO FLANGE WELD (CUT 3)
A C I /j VESSEL FLANGE TO UPPER SHELL WELD D UPPER HEAD REGION Vogtle Units 1 and 2 A 86.0 B 7.00 C 27.25 D 170.88 NOTE: ALL DIMENSIONS ARE IN INCHES Figure 2-1 Geometry of the Upper Head/Flange Region of the Vogtle Units 1 and 2 Reactor Vessels WCAP- 16142-NP February 2004 6287-NP.doc-2/1 6/04 Revision I
3-1 3 FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES The fracture evaluation was carried out using the approach suggested by Section XI Appendix G [1]. A semi-elliptic surface flaw was postulated to exist in the highest stressed region, which is at the outside surface of the closure flange. The flaw depth was assumed to encompass a range of depths into the wall thickness, and the shape was set at a length six times the depth. 3.1 STRESS INTENSITY FACTOR CALCULATIONS One of the key elements of a fracture evaluation is the determination of the driving force or stress intensity factor (K,). In most cases, the stress intensity factor for the structural integrity calculations utilized a representation of the actual stress profile rather than a linearization. The stress profile was represented by a cubic polynomial: o(x) = AO + Ajx + A2x 2 + A3 x 3 (3-1) where: x = the coordinate distance into the wall, in. 0 = stress perpendicular to the plane of the crack, ksi Ai = coefficients of the cubic fit For the surface flaw with length six times its depth, the stress intensity factor expression of Raju and Newman (Ref. 2) was used. The stress intensity factor K, can be calculated anywhere along the crack front. The point of maximum crack depth is represented by ¢ = 0, and this location was found to also be the point of maximum K, for the cases considered here. The following expression is used for calculating K, as a [unction of the angular location around the crack front (0). The units of K, are ksil/ij. K, =[-3] ZGj (a/c, a/t, tUR, A) Aj ai (3-2) The boundary correction factors Go, GI, G2 , and G3 are obtained by the procedure outlined in reference (2). The dimension "a" is the crack depth, "c" is the crack half length, "t" is the wall thickness, "R" is the inside radius, and "Q" is the flaw shape factor, which can be approximated by Q = I + 1.464 (a/c)' ". 3.2 FRACTURE TOUGHNESS Another key element in a fracture evaluation is the fracture toughness of the material. The fracture toughness has been taken directly from the reference curves of Appendix A, Section XI [I ]. In the transition temperature region, these curves can be represented by the following equations: K1, = 33.2 + 20.734 exp[O.02 (T - RTNDT)] (3-3) Ki, = 26.8 + 12.445 exp[O.0145 (T- RTNDT)] (3-4) where Kll and KIl are in ksini;. WCAP- 1642-NP February 2004 6287-N'.doc-216/04 Revision I
3-2 The upper shelf temperature regime requires utilization of a shelf toughness which is not specified in the ASME Code. A value of 200 ksikiin has been used here. This value is consistent with general practice in such evaluations, as shown for example in reference 3, which provided the background and technical basis for the development of Appendix A of Section XI. The final key element in the determination of the fracture toughness is the value of RTNDT, which is a material parameter determined from Charpy V-notch and drop-weight tests. The value of RTNDT for the closure flange region of the Vogtle units was obtained from the reactor vessel record reports [13, 14] and the certified material test reports or determined from Charpy tests and drop-weight tests [15]. The results are shown in Table 3-1. The highest value was 20'F and so this value was used for the illustrations to be discussed in Sections 4 and 5. 3.3 IRRADIATION EFFECTS Neutron irradiation has been shown to produce embrittlement which reduces the toughness properties of reactor vessel steels. The decrease in the toughness properties can be assessed by determining the shift to higher temperatures of the reference nil-ductility transition temperature, RTNDT. The location of the closure flange region is such that the irradiation levels are very low and therefore the fracture toughness is not measurably affected. a.c~e
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4 WCAP- 16142-NP February 2004 6287-NP.doc-211 6/04 Revision I
4-1 4 FLANGE INTEGRITY The first step in evaluation of the closure head/flange region is to examine the stresses. The stresses which are affected by the boltup event are the axial, or meridional stresses, which are perpendicular to the nominal plane of the closure head to flange weld. The stresses in this region during the entire heatup and cooldown process are summarized in Appendix C. The boltup is the key condition to review here, in comparison with the heatup and cooldown operation, since the flange requirement applies to boltup conditions. No other transients result in stresses in this region at low temperatures. One might suggest that the cooldown might be of similar concern, but the boltup is governing for a number of reasons: I. The heatup and cooldown transient is structured to ensure generous margins are maintained (SF = 2) for a large flaw in the irradiated beltline region, not for the unirradiated flange region.
- 2. The cooldown transient has much higher temperatures in the head region than the boltup, and
- 3. The thermal stresses caused by the cooldown transient tend to counteract the boltup stresses; cooldown thermal stresses are tensile on the inside surface and compressive on the outside surface.
Table 4-1 provides a comparison of the stresses at boltup with those at the governing time step of heatup and cooldown which is end of heatup. It is easy to see that the stresses at boltup are mostly bending, with a very small membrane stress. As the vessel is pressurized, the membrane stresses increase. These results were taken from a finite element analysis of the heatup/cooldown process, and the boltup stress alone was compared with the most limiting time step of the entire heatup/cooldown transient, which includes pressure, thermal, and boltup stresses. The relative impact of these stresses can best be addressed through a fracture evaluation. A semi-elliptic surface flaw was postulated at the outer surface of the closure head flange, and the stress intensity factor, K1, (or crack driving force) was calculated. The results are shown for cut 3 weld region in Figure 4-1. and for the cut 2 weld region in Figure 4-2. For a semi-elliptic surface flaw with depth equal to 10 percent of the wall thickness postulated in the highest stress region of the head, the following values were determined for the stress intensity factor. Boltup: K, = 24.84 ksiIii; (for a/t = 0.1 ) End of Heatup: K, = 49.21 ksitii (for a/t = 0.1) It will be useful to highlight the difference in the integrity for the head region using the two values of fracture toughness. The boltup temperature for a typical PWR is 60'F, so if RTNMrr = 201F the ASME reference toughness values are Kin = 49.0 ksi/ii; and K1 , = 79.3 ksiz/i . Using the Ki, toughness (which was the basis for the original flange requirements) it can be seen that the toughness exceeds the applied stress intensity factor for boltup for flaws of any depth in the head thickness. From Figure 4-1, the smallest margin = I- KI/K,3 = 0.39, when a/t = 0.36. For the heatup and cooldown transient, the coolant WCAP- 16142-NP February 2004 6287-NP.doc-2/1 6/04 Revision I
4-2 temperature at the governing time steps, near the end of heatup, is 557 0 F. The fracture toughness is therefore 200 ksi1iH, so again the margin is very large. Using the KI, toughness, which has now been adopted by Section XI [ I ] for P-T Curves, it can be seen that there is also a significant margin between the fracture toughness and the applied stress intensity factor, for both the boltup and the heatup cooldown transient. Another objective of the requirements in Appendix G is to assure that fracture margins are maintained to protect against service induced cracking due to environmental effects. Since the governing flaw is on the outside surface (the inside is in compression) where there are no environmental effects, there is even greater assurance of fracture margin. Therefore, it may be concluded that the integrity of the closure head/flange region is not a concern for the Vogtle units using the KI, toughness. There are two possible mechanisms of degradation for this region, thermal aging and fatigue. Effect of Fatigue. The calculated design fatigue usage for this region is less than 0. 1. so it may be concluded that flaws are unlikely to initiate in this region. la.c.e WCAP- 16142-NP February 2004 6287-NP.doc-2/1 6/04 Revision I
4-3 Table 4-1 Stress Distributions at the Closure Flange Region - Vogtle Units I and 2 Distance Boltup Stress at Cut 3 lleatup* (344 minutes) at Cut 2 (x/t) (ksi) (at p= 2 3 17 psig, t=5570 F) 0 (ID) -14.38 -15.32 0.1 -10.77 0.2 -7.83 -3.42 0.3 -5.14 0.4 -2.66 4.55 0.5 -0.26 0.6 2.16 12.15 0.7 4.72 0.8 7.54 21.76 0.9 11.24 1.0 (OD) 19.70 38.77
- With boltup stress superimposed.
Notes: Cut 3 has the highest boltup stress. Cut 2 has the highest transient stress. February 2004I Revision WCAP-l 6142-NP WCAP-16142-NP February 2004 6287.NP.doc-2/I 6/04 6287-NP.doc-2/16104 Revision I
4-4 Stress Intensity Factor vs aft for Outside Surface Flaw (Apect Ratlo=6:1) at Cut 3 C, U. C0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 08 a/i ratio (flaw depth/wall thickness) Figure 4-1 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Torus to Flange Region Weld for Vogtle Units I and 2 (stress intensity factor units are ksiiS ) WCAP- 16142-NP February 2004 6287-NP.doc-2/ 6/04 Revision I
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0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 ant ratio (flaw depthhvall thickness) Figure 4-2 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Dome to Torus Region Weld for Vogtle Units 1 and 2 (stress intensity factor units are ksijin ) WCAP-1 6142-NP February 2004 6287.NP.doc-2/16/04 Revision I
5-1 5 ARE FLANGE REQUIREMENTS NECESSARY? Using the Kic curve can support the elimination of the flange temperature requirement. This can be illustrated by examining the stress intensity factor change for a postulated flaw as the vessel is heated and pressurized after boltup, progressing up to steady state operation. The stresses at the region of interest are shown in Table 4-1, for the end of heatup, as well as boltup. Included here are the stress distributions through the wall, showing that the highest stress location for this region is the outer surface. As the vessel is pressurized, the stresses in the closure flange region gradually change from mostly bending stresses to a combination of bending and membrane stresses. The stress intensity factor, or driving force, increases for a postulated flaw at the outside surface, as the vessel is pressurized. A direct comparison between the original basis for the boltup requirement and the new Kic approach is provided in Table 5-1. This table provides calculated boltup requirements for all the designs, using a safety factor of 2, and a reference flaw depth of a/t = 0.10, which was used by Randall as the basis for the original requirement 1II]. Before discussing the table, it will be helpful to discuss the basis for the reference flaw, in light of current technology, and using the results of the Performance Demonstration Initiative. Basis for the Reference Flaw Size. Regulatory Guide 1.150 stimulated improvement in examinations of the clad to base-metal interface. The same techniques have been used for more than 10 years for reactor vessel head examinations performed from the outside surface. Capability demonstrations for the clad to base-metal interface have been conducted at the EPRI NDE Center since 1983. These demonstrations were performed initially for the belt-line region. However, similar techniques are used for both the vessel belt-line and the reactor vessel head, although the head exams are done manually. [ Iac WCAP-1 6142-NP February 2004 6287-NP.doc-2116104 Revision I
5-2
]a.c.e The flange temperature requirement. Finally, a simple illustration can be used to demonstrate clearly that no flange requirement is needed at the Vogtle units. The lower bound K1 cfracture toughness from the ASME Code is 33.2 ksi-sq-rt-in., which means that the toughness cannot be lower than that value. Study of Figures 4-1 and 4-2, for a postulated flaw of ten percent of the thickness, shows that the stress intensity factor does not exceed this value until a time step of about 300 minutes, or 5 hours into the heatup. By this time, the flange temperature is greater than 200'F, so there can be no possibility of fracture. All other locations within the flange have lower stresses, so this statement applies to the entire flange. This clearly shows that the flange requirement can be eliminated for Vogtle Units I and 2.
WCAP- 16142-NP February 2004 6287-NP.doc-21 6/04 Revision I
5-3 Table 5-1 Comparison of Various Plant Designs Boltup Requirements K, (ksi~i) T - RTNDT (F) T - RTNDt (fF) K, (ksi ) (with alt = 0.1, using KI, using K], Plant (a/t =.1) SF = 2) (alt =.10) (a/lt =.10) CE 30.0 60.0 13 68 B&W 39.4 79.8 41 100 Vogtle Units 24.9 49.8 0* 43 W 3 Loop 28.7 57.5 8 . 63 GE (CBI 251") 38.7 77.4 38 97 GE(B&W251") 48.0 96.0 56 118 GE (CE 218") 25.1 50.2 0* 43
- The calculated value of T-RTNDT is negative, so zero is used for conservatism.
WCAP- 16142-NP February 2004 6287-NP.doc-211 SO Revision I
54 Figure 5-1 Probability of Correct Rejection/Reporting (PCR) Considering Passed plus Failed Candidates, Appendix VIII Supplement 4, Detection from the Outside Surface. Reporting Criterion A' = 0.15 inch, TWE Represents Flaw Depth. WCAP- 16142-NP February 2004 6287-NP.doc-2116/04 Revision I
5-5 a,c,e Figure 5-2 Probability of Correct Rejection/Reporting (PCR) Considering Only Passed Candidates, Appendix VIII Supplement 4, Detection from the Outside Surface. Reporting Criterion A' = 0.15 inch, TWE Represents Flaw Depth. WCAP- 16142-NP February 2004 6287-NP.doc-2116/04 Revision .I
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6-1 6 SAFETY IMPLICATIONS OF THE FLANGE REQUIREMENT There are important safety implications which are associated with the flange requirement, as illustrated by Figure 6-1. The safety concern is the narrow operating window at low temperatures forced by the flange requirement. The flange requirement sets a pressure limit of 621 psi for a PWR (20 percent of hydrotest pressure). Thus, no matter how good the toughness of the vessel,-the P-T limit curve may be superceded by the flange requirement for temperatures below RTNDr + 120'F. This requirement was originally imposed to ensure the integrity of the flange region during boltup, but Section 4 has shown that this is no longer a concern. The flange requirement can cause severe operational limitations when instrument uncertainties are added to the lower temperature range limit (621 psi), for the Low Temperature Overpressure Protection system of PWRs. The minimum pressure required to cool the seals of the main coolant pumps is 325 psi, so the operating window sometimes becomes very small, as shown schematically in Figure 6-1. If the operator allows the pressure to drop below the pump seal limit, the seals could fail, causing the equivalent of a small break LOCA, a significant safety problem. Elimination of the flange requirement will significantly widen the operating window for most PWRs. An example will be provided to illustrate this situation for an operating PWR plant, Byron Unit 1. This is a forging-limited vessel at 12 EFPY, with a low leakage core, and low copper weld material in the core region. The vessel has excellent fracture toughness, which means that the flange notch is very prominent. as shown in the vessel heatup curve of Figure 6-2. As illustrated before in Figure 6-1. Byron has the LTOP setpoints significantly below the flange requirement of 621 psi, because of a relatively large instrument uncertainty. The setpoints of the two power operated relief valves arc staggered by about 16 psi to prevent a simultaneous activation. The two PORVs have different instrument uncertainties. and for conservatism the higher uncertainty is used. A similar situation exists for cooldown. as shown in Figure 6-3. Elimination of the flange requirement for the case of Byron Unit I would mean that the PORV curve could become level at 604/587 psig, which are the leading/trailing sctpoints to protect the PORV downstream piping, through the temperature range of the 350'F down to boltup at 60'F. The operatiny window between the leading PORV and the pump seal limit rises from 121 psig (446-325) to 262 psig (587-325). This change will make a significant improvement in plant safety by reducing the prohabilit of a small LOCA, and easing the burden on the operators. This is only one example of the impact of the flange requirement. Every operating PWR plant will have a different situation, but the operational safety level will certainly be generally improved by the elimination of this unnecessary requirement. The flange impact for Vogtle Unit 1, for example, is shown in Figures 64 and 6-5 [12]. WCAP-16142-NP February 2004 6287-NP.doc-2/161O4 Revision I
6-2
/'-Hetup Curve Instrument Uncertainty LTOP to 621 pi a+ .,I IOpmng Mxo#
Pump Seal Limit 325 psi RTrx+120 Temperature Figure 6-1 Illustration of the Flange Requirement and its Effect on the Operating Window for a Typical Heatup Curve WCAP- 16142-NP February 2004 6287-NP.doc-2/1 604 Revision I
6-3 LIMITING MATERLAL2 INTERMEDIATE SHELL FORGING 5P-5933 (ush Wrv. capsule daa) LIMITING ART VALUES AT 12 EFPY: .114T,'70 F 314T, 60°F 2500 tsI II I I I I II I II I I lIl l I I ___ 1 2250 X LI 1XIT XXAX fBI? __ ASlKT
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I53s lTt] I I I I I I ISNOICK 1331 3 up S0 23.6 311? I I I I I I I I 0 I I 560 100 150 .-200--250 300- 350 400 450 500 Indicated Temperature (Deg.F) I , I Figure 6-2 Illustration of the Actual Operating Window for Heatup of Byron Unit I, a Low Copper Plant at 12 EFPY WCAP-I 6142-NP February 2004 6287-NP.doc-2116/04 Revision I
6-4 LIMITING MATERIAL: INTERIJEDlATE SHELL FORGING 5P-5933 (using unr. mvpwLe data) LIMITING ART VALUES AT 12 EFPY: 1t4T, 70°F 3/4T. 60°F 250 0 ..- ..... -.- .... 11 IIi I I I I I I I I I I I I
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- l * - - - I - - l I 0 50 I 100 l50 I 2 200 2250 300 350 400 45 56o I n d i c a t e d Temp e r. iture ( De g . F Figure 6-3 Illustration of the Actual Operating Window for Cooldown of Byron Unit 1, a Low Copper Plant at 12 EFPY WCAP- 16142-NP February 2004 February 2004 6287-NP.doc-.216/04 Revision I
6-5 VOGTLE ELECTRIC GENERAniNG PLANT (VEGP) - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT I0. I I I1 I I I ii J1II - ..I... 50 100 :150 200 '250 300 350 400 450 500 ki fl asd Ts r ru z ( 'F) Figure 6-4 Illustration of the Flange Notch for Vogtic Unit 1, Heatup Curve WCAP-16142-NP February 2004 6287-NP.doc-21610-4 Revision I
6-6 VOGTLE ELECTRIC GENERATING PLANT (VEGP) -; UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT I 0 50 100 .150 200 250 300 350 400 450 W00 "dcald Tm ws a) Figure 6-5 Illustration of the Flange Notch for Vogtle Unit 1, Cooldown Curves WCAP- 16142-NP reDruary LUU4 6287-NP.doc-2/16/04 Revision I
7-1 7 REFERENCES
- 1. ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda, ASME, New York.
- 2. Raju, I. S. and Newman, J. C. Jr., "Stress Intensity Factor Influence Coefficients for Internal and External Surface Cracks in Cylindrical Vessels," Trans. ASME, Journal of Pressure Vessel Technology, Vol. 104, pp. 293-98, 1982.
- 3. Marston, T. U., ed., "Flaw Evaluation Procedures: ASME Section XI," Electric Power Research Institute Report EPRI-NP-719 SR, August 1978.
- 4. Mitchell, M. A., "RPV P-T Limits and RPV Flange Requirements; Potential Exemptions from the Requirements of 10 CFR Part 50, Appendix G." presentation to ASME Boiler and Pressure Vessel Code, Section XI, Working Group on Operating Plant Criteria, Hollywood, FL, September 10, 2002.
- 5. Nanstad, R. K., et al., PreliminaryReview of Data Regarding Chemical Compositionand Thermal Embrittlement of Reactor Vessel Steels, ORNLUNRC/LTR-95/1, Oak Ridge, TN, January 1995.
- 6. DeVan, M. J., Lowe, Jr., A. L., and Wade, S., "Evaluation of Thermally-Aged Plates, Forgings, and Submerged Arc Weld Metals," Effects of Radiation on Alaterials: 16th International
'Symposium. ASTAM STP 1175, Philadelphia, PA,1993.
- 7. Kirk, M., "Revision of AT30 Embrittlement Trend Curves," presented at the EPRI MRP/NRC PTS Re-Evaluation meeting in Rockville, MD, August 30, 2000.
- 8. Charpy Embrittlement Correlations- Status of Combined Mechanisticand StatisticalBasesfor U.S. RPV Steels (AfRP-45); PWVR MaterialsReliabiliy Program (PWVRMfRP) EPRI, Palo Alto, CA: 2001, 1000705.
- 9. ASTM E 900-02, "Standard Guide for Predicting Radiation-Induced Transition Temperature Shift for Reactor Vessel Materials, E706 (IIF)," Annual Book of ASTM Standards, Vol. 12.02.
- 10. Langer, R., et al., "A Survey of Results on Aging Experiments of Pressure Vessel Materials,"
presentation at the ATHENA Workshop, Madrid, September 2002.
- 11. Randall, N., Abstract of Comments and Staff Response to Proposed Revision to 10 CFR Part 50, Appendices G and H, Published for Comment in the Federal Register, November 14,1980.
- 12. "Vogtle Electric Generating Plant Unit I and Unit 2 - Pressure and Temperature Limits Report Rev. 1," March 20,2001.
WCAP- 16142-NP February 2004 6287-NP.doc-2/ 6/04 Revision I
7-2
- 13. "The Reactor Vessel Group Records Evaluation Program Phase 11 Final Report for the Vogtle I Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," ABB Combustion Engineering Report MISC-PENG-016, Revision 00, dated October 1995.
- 14. "The Reactor Vessel Group Records Evaluation Program Phase 11 Final Report for the Vogtle 2 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," ABB Combustion Engineering Report MISC-PENG-017, Revision 00, dated October 1995.
- 15. Laubham, TJ., "Initial RTNDT Values for Vessel Head Materials," Westinghouse Letter LTR-RCDA-03-216, dated July 28, 2003.
- 16. Regulatory Guide 1.147, Revision 13," Inservice Inspection Code Case Acceptability, ASMlE Section XI, Division 1," published June 2003.
WCAP- 16142-NP February 2004 6287-NP.doc-2/16/04 Revision I
A-I APPENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY* F. L. Becker EPRI Charlotte NC 1 ABSTRACT I]
- Presented at the Joint EC-IAEATechnical Meeting on Improvements in Inservice Inspection Effectiveness. Pettan, The Netherlands, November 2002, lo be published.
WVCAP-1 6142-NP February 2004 6287-NP.doc-21 6/04 Revision I
A-2 3 DETECTION I Ia.c.e WCAP- 16142-NP February 2004 6287-NP.doc-2/16104 Revision I
A-3 3.1 OUTSIDE SURFACE DEMONSTRATION Figure 1 Probability of Detection Performance for Passed and Passed Plus Failed Candidates forAppendix VIII Supplement 4, from the Outside Surface as a function of the flaw through wall extent (TWE). Both automated and manual techniques are included. WVCAP- 16142-NP February 2004 6287-NP.doc-2116/04 Revision I
A-X a,c,e Figure 2 POD for Inside Surface Examinations, Pass and Pass + Failed Candidates, Passed and Pass Plus Failed Candidates are included. a~c,e
+ 4
+ + 4 4 4 4
.1. 4 4
+ 4 4 1" 1* 4 WCAP- 16142-NP February 2004 6287-NP.doc-2J1 6/04 Revision I
A-5 3.2 COMBINED ID AND OD DETECTION I
]a.ce a,ce Figure 3 Probability of Detection for Automated RPV Examinations Considering Both Inside and Outside Access. Passed and Passed Plus Failed Candidates are shown.
WCAP- 16142-NP February 2004 6287-NP.doc-211 6/04 Revision I
A-6 a,c,e Figure 4 POD for Pass and Failed Candidates, Considering ID and OD Automated Demonstrations and Manual OD Demonstrations. 4 SIZING I WCAP- 161 42-NP February 2004 6287-NP.doc-2116/04 Revision I
A-7 ace Figure 5 Histogram of Depth Successful Sizing Candidate Test Scores, Appendix VIII, Supplement 4. Examinations Were Performed Both From the Inside and Outside Surfaces.
]a~c.e WCAP-1 61 42-NP February 2004 6287-NP.doc-21 W604 Revision I
A-8 0 [
] a.c.e a,c,e Figure 6 Sizing Error Surface Model a.c.e Figure 7 Plan View of Sizing Error Surface Model WCAP- 16142-NP February 2004 6287-NP.doc-2116/04 Revision I
A-9 5 ACCEPTABILITY EVALUATION II
]a.ce WCAP- 161 42-NP February 2004 6287-NP.doc-211 6/04 Revision I
A-10 Ia.c.e WCAP- 16142-NP February 2004 6287-NP.doc-2/I 6/04 Revision I
A-1l a,c,e Figure 8 Probability of Correct Sizing for Passed Candidates, Appendix VIII Supplement 4. Reporting Threshold A' = 0.15 inch. WCAP- 16142-NP February 2004 6287-NP.doc-21 6/04 Revision 1
A-12 a,c,e Figure 9 Probability of Correct Rejection/Reporting (PCR) for automated techniques, Considering Passed and Passed plus Failed Candidates, includes both inside and outside surface information. Reporting Criterion A' = 0.15 inch. 6
SUMMARY
]3.c.e 7 REFERENCES
- 1. I lawc.e WCAP- 16142-NP February 2004 6287-NP.doc-2/1 6104 Revision I
A-13
- 4. [
lace WUA--1142-INF February 2004 6287-NP.doc-2/16104 Revision I
B-1 APPENDIX B THERMAL AGING OF FERRITIC RPV STEELS AT REACTOR OPERATING TEMPERATURES [ I .cc I
]l^cc WCAP-1 6142-NP February 2004 6287-NP.doc-21 6/04 Revision I
B-2 [ la.c.e WCAP-16142-NP February 2004 6287-NP.doc-2/16104 Revision I
B-3 [ I3.Cfl WCAP-16I42-NP February 2004 6287-NP.doc-2116/04 Revision I
B4 [ T 4 4 4 4 4 4 4 4. 4 I I II la.c.e February 2004I Revision WCAP- 161 42-NP 42-NP February 2004 6287-NP.doc-2116/04 6287-NP.doc-2/16/0W Revision I
B-5
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B-6 I la.c.e WVCAP- 161 42-NP February 2004 6287-NP.doc-2/1 6/04 Revision I
c-I APPENDIX C STRESS DISTRIBUTIONS IN THE CLOSURE HEAD REGION I lIz.-.c WL J-A'-iq14Z-INr February 2004 6287-NP.doc-216104 Revision I
C-2 ace 4 + 4 4 + 4 4 + I t 1- I 4 4. I 4 4-
- I 1- 1 4 4- I I 1- I 4 4. I 4 4- I 4 4. I 4 1- 1 4 4- 1 4 I 4 4. 4 4 4- 4 4 4. 4 4 1-I 4 4. 4 4 4. 4.
I 4 4 1 -l
~VCA-1642-N Febuary200 WCAP- 16142-NP February 200.4 6287-NP.doc-2116/04 Revision I
C-3 a,c,e II
.I
.I
.I
_ I I WCAP-1 6142-NP February 2004 6287-NP.doc-211 604 Revision I
C-4 ac~ a,c,e I I WCAP- 16142-NP February 2004 6287-NP.doc-2/116/04 Revision I
C-5 a,c,e WCAP-I6142-NP February 2004 6287-NP.doc-21 S6W4 Revision I
C-6 a,c,e February 2004I Revision 16142-NP WCAP- I161 42-NP February 2004 6287-NP.doc-2116/04 6287-NP.doc-2/16104 Revision I
C-7 a,c,e I I _I _ I _
+ 4-WCAP-16142-NP February 2004 6287-NP.doc-2/16104 Revision I
D-1 APPENDIX D FLANGE INSPECTION RESULTS: VOGTLE UNITS WCAP- I 6142-NP February 2004 6287-NP.doc-2/16104 Revision I
D-2 D-1 Introduction The information below provides information about the Vogtle RPV head flange region inservice inspections (when the inspections were conducted, the extent of coverage achieved, inspection methods used, etc.). This information supports the discussion in the body of this report regarding the quality of inspections cited to support the assumed reference flaw size (page 5-1). More specifically, this evaluation demonstrates how the inspections conducted at Vogtle, Units I and 2, support the assumption of a 0.IT flaw size in the flange evaluation. D-2 Inspection History The Vogtle Unit I and 2 Closure Head to Flange welds (nominal 8.25-inch thick) were examined during the second ten-year interval to the 1989 Edition of ASME Section XI. The vessel flange examination history is shown in Table D-l. Examinations have also been performed during PSI and the first ten-year interval to earlier editions of ASMIE Section XI. Magnetic particle and ultrasonic examination of the examination volume (Figure InB-2500-5) was performed as follows: Mapnetic Particle and Ultrasonic Examinations Unit l:
- 100% of the weld length was examined during the I R7 outage (Fall 1997).
- This outage was in the first ISI period of the second ten-year ISI interval.
- Both the magnetic particle and the ultrasonic examinations had no recordable indications.
Unit 2:
- 100% of the weld length was examined during the 2R7 outage (Fall 1999).
- This outage was in the first ISI period of the second ten-year ISI interval.
- Both the magnetic particle and the ultrasonic examinations had no recordable indications.
SNC submitted a relief request (RR-4) to the NRC for limited ultrasonic examination coverage for both Vogtle I and 2. This relief request was submitted as part of the second ten-year ISI Program. This relief request was approved by a safety evaluation report (SER) to SNC in an NRC letter dated December 31, 1998. D-3 Magnetic Particle Examination Techniques In order (o detect flaws open to the outer diameter (OD) surface, a magnetic particle examination of the weld and adjacent areas as required in Figure IWB-2500-5 was performed prior to each ultrasonic examination using SNC procedure MIT-V-505. This examination procedure was written in accordance with Article 7 of the ASME Boiler and Pressure Vessel Code Section V. The acceptance standards and extent of coverage was in accordance with Section XI of the ASME Boiler and Vessel Code examination category B-A. One hundred percent of the required examination surface per figure IWB-2500-5 was achieved and no indications recorded. WCAP- 16142-NP February 2004 6287-N P.doc-2/16/04 Revision I
D-3 D4 Ultrasonic Examination Techniques In addition to the magnetic particle examination, an ultrasonic examination was performed in accordance with SNC procedure UT-V41 1. This procedure was written to comply with ASME Section XI, Appendix 1,Article 4 of ASME Section V, 1989 Edition and NRC Regulatory Guide 1.150. The ultrasonic acceptance standards were in accordance with Section XI requirements. The ultrasonic examination system was calibrated on basic calibration block 402A. In order to record and evaluate flaws throughout the volume of the weld, a distance amplitude correction curve (DAC) was generated from side drilled holes and a 2% ID notch (approx. 0.105-inch deep, excluding clad) into the base material, which is used for far surface resolution. Scanning of the component is performed at a minimum of 6 dB over the reference gain at which the DAC is established. The increase in scanning sensitivity further increases the probability of detection. The 1989 Edition of ASME Section XI, Appendix I requires that reflectors that produce a response greater than 20% DAC be investigated. In addition, the examiner is required to determine whether the indication originates from a flaw or is a geometric indication. During these examinations at both Vogtle units, the ultrasonic examination had limitations due to the flange configuration and lifting lug obstructions. The combined coverage was calculated to be approximately 68%. D-5 Summary The magnetic particle surface examination will detect flaws open to the surface. The ultrasonic examinations will detect flaws throughout the volume. The ultrasonic examinations are conducted with scanning sensitivities (at least 2X or +6 dB) over calibration. The recording requirements for ultrasonic examination are extremely low (20% DAC). The ultrasonic examinations were calibrated on a 2% notch from calibration block 402A into the ferriiic base material. It is expected that an ultrasonic response from the 2% notch would be extremely sensili.C when compared to a 0.1T(10%), flaw, which is of concern. A high percentage of coverage (approximately 90%) from the head side was obtained with the ultrasonic examination. A high percentage of coverage (100%) was obtained with the magnetic particle examination. There were no indications recorded with either the magnetic or ultrasonic examinations for either the Unit I or Unit 2 head to flange weld. WCAP-16142-NP February 2004 6287-NP.doc-2/16104 Revision I
D-4 D-6 Conclusion The probability of detection for flaws on the high stress region of the outer surface of the closure head is very high due to the magnetic particle examination being performed with no limitations. Based on the ability of the ultrasonic system to detect ID surface reflectors (2% ID notch), there is very high probability that flaws 0.1 T(10%) will be detected for the accessible volumes. Although the ID surface was not fully examined in those areas where the RPV lifting lugs are located, a significant amount of the outer 25% of the examination area was scanned in all four directions by ultrasonic examination and therefore, significant through-wall indications would have been recorded. WCAP- 16142-NP February 2004 6287-NP.doc-2/16/04 Revision I
- D-5 Table D-1 Reactor Vessel Flange Examination History Component Description Examination Sensitivity Coverage Results Comments/Schedule Flange to head weld Surface examination The surface technique is Achieved 100% coverage No recordable This examination was (I 1201 -V6-001 - (magnetic particle) capable of detecting for the surface indications for performed twice during the W02) and indications with a major examination. either the surface Unit's commercial dimension of 1/16th of an or ultrasonic operation, last performed in Ultrasonic exams inch. The volumetric examinations. October 1997 (I1R7). The using 0,45, 60 and examination was limited next examination will be 70 degree scans. The sensitivity of the to approximately 65% due performed in the third U Examinations have ultrasonic exams is based to flange configuration interval (2008).
-Unt I been performed to on signal responses from and 3 integrally mounted Reactor Vessel NRC RG 1.150. calibration block side- lifting lugs, (see NRC Head drilled holes and ID notch. approved Relief Request RR-4).
Flange Visual examination Technique requires - Entire flange area on head There have been This exam is performed lighting and distance and vessel is examined, no recordable each refueling outage. sufficient to detect however, the defect size indications that scratches or pits 0.001 to criteria are specifically have required 0.003 of an inch. applied to the o-ring repair. seating surface areas. Flange ligament Ultrasonic' The sensitivity of the- Achieved 100% coverage No recordable Ligament exams divided (I1201-V6-00I-L0I examinationfof the ultrasonic exam is based for the Volumetric indications. into stud iegions (54) and through L54) threads inthe flange on signal responses from examination. spread across the ISI and a I-inch annular calibration block side- periods. The last exams volume around the drilled holes. performed March 2002 flange stud holes. (I1R10). The remaining third scheduled for March 2005 (IR12).l Unit I Flange to shell weld Mechanized The sensitivity of the Achieved 100% coverage Code acceptable This examination was Reactor Vessel (I1201-V6-001- Ultrasonic IDexams ultrasonic exam is based for the Volumetric indications W03) (tool) using 0.45, 60 on signal responses from performed once during the examination using recorded. Unit's commercial and 70 degree scans calibration block side- combination of Shell side operation in March 1996 and manual exams drilled holes and ID notch. and flange face (I1R6). The next from the flange face examinations. Expect to examination is scheduled with a "sled" achieve essentially 100% for September 2006 (1R13). apparatus. with Appendix VilI VEGP plans to submit Examinations have techniques. relief to use Appendix Vill been performed to qualified techniques in lieu NRC RG 1.150. of Section V and RG 1.150. WCAP- 16142-NP February 2004 6287-P.do-21 6/04 Revision I
D-6 Table D-1 Reactor Vessel Flange Examination History (cont.) Component Description Examination Sensitivity Coverage Results Comments/Shedlule Flange io head weld Surface examination Thc surface technique Achieved 100% No recordable This examination was (21201-V6-001- (magnetic particle) is capable of detecting coverage for the indications for either performed twice during the W02) and indications with a surface examination. the surface or Unit's commercial major dimension of The volumetrie ultrasonic operation, last performed in Ultrasonic exams 1/16th of an inch. examination was examinations. September 1999 (2R7). using 0, 45. 60 and 70 The sensitivity of the limited to The next examination will degree scans. ultrasonic exams is approximately 65% be performed in the third Examinations have ultasdonsicgeams is appoximatgel6% e interval (2008). been performed to baeonigldutofne Ui2NRC RG .150. responses from configuration and 3 Unit2 NC R 1.50. calibration block side- integrally mounted Reactor Vessel Head drilled holes and ID lifting lugs, (see NRC notch. approved Relief Request RR-4). Flange Visual examination Technique requires Entire flange area on There have been no This exam is performed lighting and distance head and vessel is recordable indications each refueling outage. sufficient to detect examined, however, that have required scratches or pits 0.001 the defect size criteria repair. to 0.003 of an inch. are specifically applied to the o-ring seating surface areas. Flange ligament Ultrasonic The sensitivity of the Achieved 100% No recordable Ligament exams divided (21201-V6-001-LOI examination of the ultrasonic exam is coverage for the indications. into stud regions (54) and through L54) threads in the flange based on signal Volumetric spread across the ISI Unit 2 and a I-inch annular responses from examination. periods. The last exams Reactor Vessel volume around the calibration block side- performed October 2002 flange stud holes. drilled holes. (2R9). The remaining third scheduled for September __________________ 2005 (2RI 1). _C. P 6 42 N Febr.ary_2004 WCAP-16142-NP February 2004 6287-P.doc-2/16/04 Revision I
D-7 Table D-l Reactor Vessel Flange Examination History (cont.) Component Description Examination Sensitivity Coverage Results Comments/Schedule Flange to shell weld Mechanized The sensitivity of the Achieved 100% Code acceptable This examination was (21201 -V6-00 I - Ultrasonic ID exams ultrasonic exam is coverage for the indications recorded. performed once during the W03) (tool) using 0.45, 60 based on signal Volumetric Unit's commercial and 70 degree scans responses from examination using operation in March 1998 and manual exams calibration block side- combination of Shell (2R6). The next from the flange face drilled holes and ID side and flange face examination is scheduled with a "sled" notch. examinations. Expect for March 2007 (2R12). apparatus. to achieve essentially VEGP plans to submit Examinations have 100% with Appendix relief to use Appendix Vill been performed to Vill techniques. qualified techniquesin lieu NRC RG 1.150. of Section V and RG 1.150. WCAP- 161 42-NP February 2004 6287-P.doc-2/16/04 Revision I
Enclosure 7 Vogtle Electric Generating Plant Units I and 2 TS, Bases, and PTLR Amendment
Enclosure 7 Vogtle Electric Generating Plant Request to Revise Technical Specifications and Pressure and Temperature Limits Report TS. Bases, and PTLR Amendment Proposed Changes The proposed changes to the Technical Specifications (TS) are as follows:
- Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), will be revised to reference to the NRC-approved methodology for developing P/T limits and COPS setpoints (WCAP-14040-A, Rev. 4) and the methodology used to justify eliminating the reactor vessel closure head/vessel flange requirements (WCAP-16142-P, Revision 1).
- Section 3.4.12, Cold Overpressure Protection Systems (COPS), will be revised to change the RCS vent size in LCO 3.4.12b from 2.14 square inches to 1.5 square inches.
The proposed changes to the VEGP Unit 1 and Unit 2 PTLRs are consistent with the revised P/T limits and COPS setpoints, and the information used to develop the revised P/T limits and COPS setpoints. Tables 1 and 2 below provide a cross-reference of the current PTLR Figure and Table number to the revised PTLR Figure and Table number, where applicable. Table 1- Changes to VEGP Unit I PTLR Current Vogtle Unit 1 PTLR Revised Vogtle Unit I PTLR Figure Number Figure Number 2.1-1 2-1 2.1-2 2-2 2.2-1 3-1 Current Vogtle Unit 1 PTLR Revised Vogtle Unit 1 PTLR Table Number Table Number 2.1-1 2-1 2.1-2 2-2 2.2-1 3-1 3.0-1 5-1 3.0-2 5-2 3.0-3 5-3 3.0-4 5-4 3.0-5 5-7 3.0-6 5-5 and 5-6 3.0-7 5-4 3.0-8 Deleted 3.0-9 5-8 3.0-10 5-9 5.0-1 Deleted E7-1
Enclosure 7 Vogtle Electric Generating Plant Request to Revise Technical Specifications and Pressure and Temperature Limits Report TS, Bases, and PTLR Amendment Table 2- Changes to VEGP Unit 2 PTLR Current Vogtle Unit 2 PTLR Revised Vogtle Unit 2 PTLR Figure Number Figure Number 2.1-1 2-1 2.1-2 2-2 2.2-1 3-1 Current Vogtle Unit 2 PTLR Revised Vogtle Unit 2 PTLR Table Number Table Number 2.1-1 2-1 2.1-2 2-2 2.2-1 3-1 3.0-1 5-1 3.0-2 5-2 3.0-3 5-3 3.0-4 5-4 3.0-5 5-7 3.0-6 5-5 and 5-6 3.0-7 5-4 3.0-8 5-8 3.0-9 5-9 Basis for Proposed Changes One of the proposed changes to the TS involves revising Section 3.4.12, Cold Overpressure Protection Systems (COPS), to change the RCS vent size in LCO 3.4.12 b. The RCS vent size was calculated based on the revised Appendix G limits developed utilizing ASME Code Case N-640 and without the reactor vessel flange temperature requirement. The revised RCS vent size is capable of mitigating the limiting cold overpressure transient. The other proposed changes to the TS involve revising the references in Section 5.6.6 to reflect the NRC-approved methodology for developing P/T limits and COPS setpoints (WCAP-14040-A, Rev. 4) and the methodology used to justify eliminating the reactor vessel closure head/vessel flange requirements (WCAP-16142-P, Revision 1). The PTLR currently contains the pressure and temperature (PAID) limits, including heatup and cooldown rates, and the nominal power operated relief valve (PORV) setpoints for cold overpressure protection. The concept of the PTLR is that the limits contained in it, which are also plant-specific and vary with vessel fluence, can be revised without prior NRC approval provided that they are calculated using an NRC-approved methodology. The revised VEGP Unit I and 2 PTLRs are consistent with NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." The surveillance capsule withdrawal schedule is contained in UFSAR Table 5.3.1-8 for Unit I and UFSAR Table 5.3.1-9 for Unit 2. The methodology for determining the limits specified in the PTLR is discussed in Enclosure 1. E7-2
Enclosure 8 Vogtle Electric Generating Plant Units I and 2 Significant Hazards Consideration Evaluation
Enclosure 8 Vogtle Electric Generating Plant Request to Revise Technical Specifications and Pressure and Temperature Limits Report Significant Hazard Consideration Evaluation Proposed Changes Southern Nuclear Operating Company (SNC) proposes to revise the Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS). The proposed changes would revise Section 3.4.12 to change the RCS vent size in LCO 3.4.12 b and would revise Section 5.6.6 to incorporate references to WCAP-14040-A, Rev. 4 and WCAP-16142-P, Revision 1. The revised VEGP Unit I and 2 PTLRs are consistent with NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." The following is a detailed description of the proposed changes. Section 3.4.12, Cold Overpressure Protection Systems (COPS), will be revised to change the RCS vent size in LCO 3.4.12 b from 2.14 square inches to 1.5 square inches. Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), will be revised to reference to the NRC-approved methodology for developing P/T limits and COPS setpoints (WCAP-14040-A, Rev. 4) and the methodology used to justify eliminating the reactor vessel closure head/vessel flange requirements (WCAP-16 142-P, Revision 1). Evaluation
- 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
No. The proposed changes to the Technical Specifications and PTLRs do not affect any plant equipment, test methods, or plant operation, and are not initiators of any analyzed accident sequence. Operation in accordance with the proposed TS will ensure that all analyzed accidents will continue to be mitigated by the SSCs as previously analyzed. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
No. The proposed changes do not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. The changes to the P-T limits and COPS setpoints will ensure that appropriate fracture toughness margins are maintained to protect against reactor vessel failure during both normal and low temperature operation. The changes to the P-T limits and COPS setpoints are consistent with the methodology approved by the NRC in WCAP-14040, Rev. 4. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Do the proposed changes involve a significant reduction in a margin of safety?
No. The proposed changes will not adversely affect the operation of plant equipment or the function of any equipment assumed in the accident analysis. The utilization of ASME Code Case N-640 maintains the relative margin of safety commensurate with that which existed at the time that ASME B&PV Code, Section XI, Appendix G was approved in 1974 and will ensure an acceptable margin of safety. Therefore, the proposed changes do no involve a significant reduction in any margin of safety. E8- I
Enclosure 9 Vogtle Electric Generating Plant Units I and 2 Pen and Ink Changes
Technical Specification Changes COPS 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Cold Overpressure Protection Systems (COPS) LCO 3.4.12 A COPS shall be OPERABLE with all safety injection pumps incapable of injecting into the RCS and the accumulators isolated and either a or b below.
- a. Two RCS relief valves, as follows:
- 1. Two power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, or
- 2. Two residual heat removal (RHR) suction relief valves with setpoints.
- Ž440 psig and
- 460 psig, or
- 3. One PORV with a lift setting within the limits specified in the PTLR and one RHR suction relief valve-with a setpoint within specified limits.
- b. The RCS depressurized and an RCS vent of Ir uare inches (based on an equivalent length of 10 feet of pipe).
APPLICABILITY: MODE 4, MODE 5, MODE 6 when the reactor vessel head is on.
----- NOTE-
- 1. Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR.
- 2. The safety injection pumps are not required to be incapable of injecting into the RCS until 4 hours after entering MODE 4 from MODE 3 provided the temperature of one or more RCS cold legs has not decreased below 3250F.
Vogtle Units 1 and 2 3.4.12-1 Amendment No. 96 (Uniti ) Amendment No. 74 (Unit 2)
L Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report (COLR) (continued)
- d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant-System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a. RCS pressure and temperature limits torheatup, cooldown, operation, criticality, and hydrostatic testing as well as heatup and coutdown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3 "RCS Pressure and Temperature (PIT) Limits"
- b. The power operated relief valve lift settings required to support the Cold Overpressure Protection Systems (COPS) shall be established and documented in the PTLR for the following:
LCO 3.4.12 "Cold Overpressure Protection Systems" C. Thp RCSpr sure and temperature limits for U i 1 shall be those envious reviewed and aproved by the N_5Z in Amendment No.,,87 to Facili perating Licen IsNPF-68. The RS pressure and tenerature limi for Unit 2 shall those previou reviewed and appr d by the RC in Amendme No. 65 to Facil Operating License F-81. The acceptability of e P/T and COP imits are document in NRC letter "Vogtle Elec Generating Pi Units 1 and 2 - Ac ptance for Referenci of Pressure Te erature Limits Repr(" February 12, 1996. ecifically, the Ii s and methodology e described in th folio g documents:
- 1. Amendment o. 87 to Facility Op ting License No PF-68, Vogtle El c Generating Plan1Unit 1, June 8, 1 5.
- 2. Ame ment No. 65 to Faci Operating Lice e No. NPF-81, Vo le Electfic Generatin Plant, Unit 2, Ju 8, 1995. 2 (continued)
Vogtle Units I and 2 5.64 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2)
Insert 'The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
- 2. WCAP-16142-P C"Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units I and 2."
Reporting Requirements _ --- -. 5.6 --- -5.6-Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) Letty fromC. I. Grimes, NRC R. A. Newton, Yestinghou se lctric irporation, "Accep nce for Referen g of Top I
,tepo CAP-14040, Re -iion1, 'Meth ogy Use o Deqelop Cold Overpre ure Mitigating System Setp nts and R keatup and Cooldo mifCirves Oc-tb-1 , 1995. /
Letter from C. McCoy, Geo a Power C pany, to U.S. Nucle egulatory C mission, A ntion: Docu ent - Control Dsk, OVogtle El riG Genera*g Plant, Pre ure and Tempe ture Limits Reort,' Enclos es 1 and 2, nuary 26, 1996.
- d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement. I 5.6.8 PAM Report When a Report is required by Condition G or K of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation,' a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. (continued) Vogtle Units 1 and 2 5.6-5 Amendment No. 117 (Unit 1) Amendment No. 95 (Unit 2)
Bases Changes RCS P/T Limits B 3.4.3 _1 B-3A4REACTOR-COOLANT SYSTEM (RCS) B 3:4.3 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND All components of theRCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. The PTLR contains PIT limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design' characteristics and operating functions. 10 CFR 50, Appendix G (Ref. 1), requires the establishment of PIT limits for specific material fracture toughness requirements of the RCPB materials. Reference I requires an adequate margin to brittle
-failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section)11 Appendix G (Ref. 2).
The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTmn) as exposure to neutron fluence increases. (continued) Vogtle Units I and 2 B 3.4.3-1 Revision No. 0
RCS PIT Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued) Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time. Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, SectionX1 Appendix G.
- 3. ASTM E 185-82, July 1982.
- 4. 10 CFR 50, Appendix H.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
(continued) Vogtle Units 1 and 2 B 3.4.3-7 Revision No. 0
RCS PIT Limits
- B 3.4.3 BASES c REFERENCES 6. ASME, Boiler and Pressure Vessel Code, Section Xl, (continued) Appendix E.
- 7. WCAP-14040, Revision/, I D A
VogUe Units 1 and 2 B 3A.3-8 Revision No. 0
COPS B 3.4.12 BASES (continued) APPLICABLE
- Safety analyses (Ref. 4) demonstrate that the reactor vessel SAFETY ANALYSES is adequately protected against exceeding the Reference 1 PIT limits.
In MODES 1, 2, and 3, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference I limits. In MODE 4 and below, overpressure prevention falls to two OPERABLE RCS relief valvesor to a depressurized RCS-and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability. The actual temperature at which the pressure in the PIT limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the COPS must be re-evaluated to ensure its functional requirements can still be met using the RCS relief valve method or the depressurized and. vented RCS condition. The PTLR contains the acceptance limits that define the COPS requirements. Any change to the RCS must be evaluated against the Reference 4 analyses to determine the impact of the change on the
- COPS acceptance limits. -
Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input tran sients Mass Input Type Transients
- a. Inadvertent safety injection; or.
- b. Charginglletdown flow mismatch.
Heat Input Type Transients
- c: Reactor coolant pump (RCP) startup with temperature asymmetryQrmn-the between the RCS and steam generators. 6 (continued)
Vogtle Units 1 and 2 B 3.4.12-4 Revision No. 0
COPS B 3.4.12 BASES APPLICABLE hetHayeIcniud , SAFETY ANALYSES ~---- The following are required during the COPS MODES to ensure that mass and heat input transients do not occur, which either of the COPS overpressure protection means cannot handle:
- a. Rendering both safety injection pumps incapable of injection;
- b. Deactivating the accumulator discharge isolation valves in their closed positions; and
- c. Disallowing start of an RCP if secondary temperature is more than 500 F above primary temperature in any one loop. With no reactor coolant pump running, this value is reduced to 250 F at an RCS temperature of 3500 F and varies linearly to 500 F at an RCS temperature of 2000 F. LCO 3.4.6, *RCS Loops-MODE 4," and LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," provide this protection.
The Reference 4 analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when both centrifugal charging pumps are actuated. Thus, the LCO requires both safety injection pumps to be incapable of injecting into the RCS during the COPS MODES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient caused by accumulator injection when RCS temperature is low, the LCO also requires accumulator isolation when accumulator pressure is great n or equal to the maximum RCS A pressure for the existing RC 6Jd leg temperature allowed in the PTLR. The isolated accumutots must have their discharge valves closed and the valve power supply breakers fixed in their open positions. PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limits shown in the PTLR. The setpoints are derived by analyses that model the performance of the COPS, assuming
- J (continued)
Vogtle Units 1 and 2 B 3.4.12-5 Revision No. 0
COPS B 3.4.12 BASES APPLICABLE PORV Performance (continued) SAFETY ANALYSES the mass injection transient of two centrifugal charging pumps and the positive displacement pump injecting into the RCS, and the heat injection .transient of starting an RCR with the RCS 500 F colder than the secondary coolant. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.
-__NOTE-----
Although the positive displacement pump (PDP) was replaced with the normal charging pump (NCP), the current mass injection transient analysis assumes two centrifugal charging pumps and the positive displacement pump. Westinghouse performed an evaluation of the effect of replacing the PDP with the NCP and obtained acceptable results without reanalysis of the mass injection transient. Reference A Westinghouse letter, GP-1 6U8 from J. L Tain to J. B. Beasley, Jr., Y dated August 13, 1998, COPS PORV Setpoint for New Charging Pumps - r The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the COPS analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3,
'RCS Pressure and Temperature (PIT) Limits," discuss these examinations.
The PORVs are considered active components. Thus, the failure of
- -one PORV-is assumed to represent the worst case, single active failure.
RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints like the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 440 psig and 460 psig (Ref. 9) will pass flow greater than that required for the limiting COPS transient while maintaining RCS pressure less than the P/T limit curve. (continued) Vogtle Units 1 and 2 B 3.4.12-6 Rev. 1-3/01
Cops B 3.4.12 BASES' APPLICABLE RHR Suction Relief Valve Performance (continued) SAFETY ANALYSES As the RCS P/T limits are decreased to reflect the loss of toughness in the reactor vessel materials due to neutron embrittlement, the RHR suction relief valves must be analyzed to still accommodate the design basis transients for COPS. The RHR suction relief valves.are considered active components. Thus, the failure of one valve is assumed to represent the worst case single active failure. RCS Vent Performance With the RCS depressurized, analyses show a vent size of 2X14 square inches (based on an equivalent length of.i10 feet of pipe; i.e., a vent capable of relievin 0 gpm waterflow at 4ix9 is capable of miaigating the allowed COPS overpressure transient. The Q5/ capacity of a vent this size is greater than the flow of the limiting transient for the COPS configuration, with both safety injection pumps incapable of injecting into the RCS, maintaining RCS pressure less than the maximum pressure on the PIT limit curve. The RCS vent size will be re-evaluated for compliance each time the PIT limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. The COPS satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii). LCO This LCO requires that the COPS is OPERABLE. The COPS is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient. To limit the coolant input capability, the LCO requires both safety injection pumps to be incapable of injecting into the RCS and all accumulator discharge isolation valves closed and immobilized when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. (continued) Vogtle Units 1 and 2 B 3.4.12-7 Rev. 2-1 0101
COPS B 3.4.12 BASES LCO The elements of the LCO that provide low temperature (continued) overpressure mitigation through pressure relief are:
- a. Two RCS relief valves, as follows:
-'1.' -'Two OPERABLE PORVs;-or A PORV is OPERABLE for the COPS when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint, and motive power is available to the two valves and their control circuits. The PORVs (PV-455A and PV-456A) are powered from 125 V MCCs 1/2AD1M and 1/2BDlM, respectively. The PORVs are to be considered OPERABLE whenever these MCCs are available to supply power.
- 2. Two OPERABLE RHR suction relief valves; or An RHR suction relief valve is OPERABLE for the COPS when its RHR suction isolation valve and its RHR suction valve are open, its setpoint is at or between 440 psig and 460 psig, and testing has proven its ability to open at this setpoint.
- 3. One OPERABLE PORV and one OPERABLE RHR suction relief valve; or
- b. A depressurized RCS and an RCS vent.
An RCS vent is OPERABLE when open with an area of
/,,.4ssquare inches (based on an equivalent length of 10 feet
; 6f'pipej i.e., capable of relievin gRpm at,4Zig).
Each of these methods of overpressure prevention is capable of mitigating the liming COPS transient. APPUCABILITY This LCO is applicable in MODE 4, in MODE 5, and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 PIT limits. When the reactor vessel head is off, overpressurization cannot occt7) LCO 3.4.3 provides the operational P/T limits for al MODES. LCO 3.4.10, Pressurizer Safety Valves," requires the (continued) Vogtle Units 1 and 2 B 3.4.12-8 Rev. 1-1/00
COPS B 3.4.12 .- BASES ACTIONS E.1 (continued) The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 7). Thus, with one of the two RCS relief valves inoperable in MODE 5 or in MODE 6 with
-the head on, the-Completion Time to restore two valves to OPERABLE status is 24 hours.
The Completion Time represents a reasonable lime to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events. F.1 The RCS must be depressurized and a vent must be established within 12 hours when:
- a. Both required RCS relief valves are inoperable; or
- b. A Required Action and associated Completion Time of Condition A, C, D, or E is not met; or
- c. The COPS is inoperable 'for any' reason other than Condition A, B, C, D, or E.
The vent must be sized 4s~quare inches (based on an equivalent length of 10 feet of pipe) to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This actionris needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel. UL4iZ, The Completion Time considers the time required to place the planTin this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements. (continued) Vogtle Units 1 and 2 B 3.4.12-11 Rev. 1-9199 l
COPS B 3.4.12 BASES SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, both safety injection pumps are verified incapable of injecting into the RCS, and the accumulator discharge isolation valves are verified closed and locked out. The safety injection pumps are rendered incapable of injecting into the RCS through at least two independent means such that a single failure or single action will not result in an injection into the RCS. The Frequency of within 4 hours after initial entry into MODE 4 from MODE 3 and prior to RCS cold leg temperature decreasing below 3250 F (for the safety injection pumps) and 12 hours thereafter (for the safety injection pumps and accumulators) is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment. SR 3.4.12.3 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and by testing it in accordance with the Inservice Testing Program. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO. For Train A, the RHR suction relief valve is PSV-8708A and the suction isolation valves are HV-8701A and B. For Train B, the RHR suction relief valve is PSV-8708B and the suction isolation valves are HV-8702A and B. The RHR suction valves are verified to be opened every 12 hours. The Frequency is considered adequate in view of other administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction isolation valves remain open. The ASME Code, Section Xl (Ref. 8), test per Inservice Testing Program verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint. SR 3.4.12.4 The RCS vent of >,X24 square inches (based on an equivalent length of 10 feet of pipe) is proven OPERABLE by verifying its open condition either: (continued) Vogtle Units 1 and 2 B 3.4.12-12 Rev. 2-9/99 l
COPS B 3.4.12 BASES.
- SURVEILLANCE SR 3.4.12.4 (cbntinued)-
REQUIREMENTS
- a. Once every 12 hours for a valve that cannot be locked.
- b. Once every 31'days for a valve that is locked, sealed, or secured in position. A removed pressurizer safety valve fits this category.
The passive vent arrangement must only be open to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.1 2b.
,4,(
SR 3.4.12.5 The PORV block valve must be verified open every 72 hours to provide the flow path for each required PORV to perform its function when actuated. The valve must be remotely verified open in the main control room. This Surveillance is performed if the PORV satisfies the LCO. The block valve is a remotely controlled, iotr operated valve. The ower to the valve operator is not requiredemoved, and the manual operator is not require ocked in the inactive position. Thus, the block valve can be clsed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation. The 72 hour Frequency is considered adequate in view of other administrative controls available to the opierator in the control room, such as valve position indication, that verify that the PORV block valve remains open. SR 3.4.12.6 Performance of a COT is requidithin2 hurs after decreasing RCS temperature to
- 3500 F and every 31 days on each required PORV to verify and, as necessary, adjust- lift setpoint. The COT will verify the setpoint is within th3 Palowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required.
A Note has been added indicating that this SR is required to be performed 12 hours after decreasing RCS cold leg temperature to S 3500F. The 12 hours considers the unlikelihood of a low temperature overpressure event during this time. (continued) Vogtle Units 1 and 2 8 3.4.12-13 Rev. 1-9/99 l
W,/
/ /0/l /c~-I VOG ELECTRIC GENERATING ;9f - UNIT 1 PRESSURE D TEMPERATURES lITS REPORT by: 01
lcc- 0!57 L-7 I 7'1L PF77 gv/y/"0/? z VOGTLE tiC /p1ANT - UNIT 2 LIMITS REPORT Prepared vko 3-1q-01 Revii Iby:
Enclosure 10 Vogtle Electric Generating Plant Units 1 and 2 Final TS, Bases, and PTLR Changes
COPS 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Cold Overpressure Protection Systems (COPS) LCO 3.4.12 A COPS shall be OPERABLE with all safety injection pumps incapable of injecting into the RCS and the accumulators isolated and either a or b below.
- a. Two RCS relief valves, as follows:
- 1. Two power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, or
- 2. Two residual heat removal (RHR) suction relief valves with setpoints
Ž 440 psig and s 460 psig, or
- 3. One PORV with a lift setting within the limits specified in the PTLR and one RHR suction relief valve with a setpoint within specified limits.
- b. The RCS depressurized and an RCS vent of Ž 1.5 square inches (based on an equivalent length of 10 feet of pipe).
APPLICABILITY: MODE 4, MODE 5, MODE 6 when the reactor vessel head is on.
--------------------------------------------- rNOT E------------------------- .
- 1. Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR.
- 2. The safety injection pumps are not required to be incapable of injecting into the RCS until 4 hours after entering MODE 4 from MODE 3 provided the temperature of one or more RCS cold legs has not decreased below 3250F.
Vogtle Units 1 and 2 3.4.12-1 Amendment No. (Uniti ) Amendment No. (Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report (COLR) (continued)
- d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a. RCS pressure and temperature limits for heatup, cooldown, operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3 "RCS Pressure and Temperature (PIT) Limits"
- b. The power operated relief valve lift settings required to support the Cold Overpressure Protection Systems (COPS) shall be established and documented in the PTLR for the following:
LCO 3.4.12 "Cold Overpressure Protection Systems"
- c. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
- 2. WCAP-16142-P, Rev. 1, 'Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Vogtle Units 1 and 2."
(continued) Vogtle Units 1 and 2 5.6-4 Amendment No. (Unit 1) Amendment No. (Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) I
- d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement. 5.6.8 PAM Report When a Report is required by Condition G or K of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. (continued) Vogtle Units 1 and 2 5.6-5 Amendment No. (Unit 1) Amendment No. (Unit 2)
RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.3 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. The PTLR contains P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions. 10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section Xl, Appendix G (Ref. 2). The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases. (continued) Vogtle Units 1 and 2 B 3.4.3-1
RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued) Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time. Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement. REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, Section Xl, I Appendix G.
- 3. ASTM E 185-82, July 1982.
- 4. 10 CFR 50, Appendix H.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
(continued) Vogtle Units 1 and 2 B 3.4.3-7
RCS P/T Limits B 3.4.3 BASES REFERENCES 6. ASME, Boiler and Pressure Vessel Code, Section Xl, (continued) Appendix E.
- 7. WCAP-14040-A, Revision 4. I Vogtle Units 1 and 2 B 3.4.3-8
COPS B 3.4.12 BASES (continued) APPLICABLE Safety analyses (Ref. 4) demonstrate that the reactor vessel SAFETY ANALYSES is adequately protected against exceeding the Reference 1 P/T limits. In MODES 1, 2, and 3, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference 1 limits. In MODE 4 and below, overpressure prevention falls to two OPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability. The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the COPS must be re-evaluated to ensure its functional requirements can still be met using the RCS relief valve method or the depressurized and vented RCS condition. The PTLR contains the acceptance limits that define the COPS requirements. Any change to the RCS must be evaluated against the Reference 4 analyses to determine the impact of the change on the COPS acceptance limits. Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients as discussed below. I Mass Input Type Transients
- a. Inadvertent safety injection; or
- b. ChargingAetdown flow mismatch.
Heat Input Type Transients a.. Reactor coolant pump (RCP) startup with temperature asymmetry between the RCS and steam generators. I (continued) Vogtle Units 1 and 2 B 3.4.12-4
COPS B 3.4.12 BASES APPLICABLE The following are required during the COPS MODES to ensure that SAFETY ANALYSES mass and heat input transients do not occur, which either of the COPS (continued) overpressure protection means cannot handle:
- a. Rendering both safety injection pumps incapable of injection;
- b. Deactivating the accumulator discharge isolation valves in their closed positions; and
- c. Disallowing start of an RCP if secondary temperature is more than 500 F above primary temperature in any one loop. With no reactor coolant pump running, this value is reduced to 250 F at an RCS temperature of 3500 F and varies linearly to 500 F at an RCS temperature of 2000F. LCO 3.4.6, "RCS Loops-MODE 4," and LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," provide this protection.
The Reference 4 analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when both centrifugal charging pumps are actuated. Thus, the LCO requires both safety injection pumps to be incapable of injecting into the RCS during the COPS MODES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient caused by accumulator injection when RCS temperature is low, the LCO also requires accumulator isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. The isolated accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions. PORV Performance' The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limits shown in the PTLR. The setpoints are derived by analyses that model the performance of the COPS, assuming (continued) Vogtle Units 1 and 2 8 3.4.12-5
COPS B 3.4.12 BASES APPLICABLE PORV Performance (continued) SAFETY ANALYSES the mass injection transient of two centrifugal charging pumps and the positive displacement pump injecting into the RCS, and the heat injection transient of starting an RCP with the RCS 500 F colder than the secondary coolant. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.
------------------------------------------- NOTE----------------------------------------
Although the positive displacement pump (PDP) was replaced with the normal charging pump (NCP), the current mass injection transient analysis assumes two centrifugal charging pumps and the positive displacement pump. Westinghouse performed an evaluation of the effect of replacing the PDP with the NCP and obtained acceptable results without reanalysis of the mass injection transient. Reference Westinghouse letter, GP-1 6838 from J. L. Tain to J. B. Beasley, Jr., dated August 13, 1998, COPS PORV Setpoint for New Charging Pump. The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the COPS analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," discuss these examinations. The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure. RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints like the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 440 psig and 460 psig (Ref. 9) will pass flow greater than that required for the limiting COPS transient while maintaining RCS pressure less than the PIT limit curve. (continued) Vogtle Units 1 and 2 B 3.4.12-6
COPS B 3.4.12 BASES APPLICABLE RHR Suction Relief Valve Performance (continued) SAFETY ANALYSES As the RCS P/T limits are decreased to reflect the loss of toughness in the reactor vessel materials due to neutron embrittlement, the.RHR suction relief valves must be analyzed to still accommodate the design basis transients for COPS. The RHR suction relief valves are considered active components. Thus, the failure of one valve is assumed to represent the worst case single active failure. RCS Vent Performance With the RCS depressurized, analyses show a vent size of 1.5 square inches (based on an equivalent length of 10 feet of pipe, i.e., a vent capable of relieving 685 gpm waterflow at 722 psig) is capable of mitigating the allowed COPS overpressure transient. The capacity of a vent this size is greater than the flow of the limiting transient for the COPS configuration, with both safety injection pumps incapable of injecting into the RCS, maintaining RCS pressure less than the maximum pressure on the P/T limit curve. The RCS vent size will be re-evaluated for compliance each time the P[T limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. The COPS satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii). LCO This LCO requires that the COPS is OPERABLE. The COPS is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient. To limit the coolant input capability, the LCO requires both safety injection pumps to be incapable of injecting into the RCS and all accumulator discharge isolation valves closed and immobilized when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. (continued) Vogtle Units 1 and 2 B 3.4.12-7
COPS B 3.4.12 BASES LCO The elements of the LCO that provide low temperature (continued) overpressure mitigation through pressure relief are:
- a. Two RCS relief valves, as follows:
- 1. Two OPERABLE PORVs; or A PORV is OPERABLE for the COPS when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint, and motive power is available to the two valves and their control circuits. The PORVs (PV-455A and PV-456A) are powered from 125 V MCCs 1/2AD1M and 1/2BD1M, respectively. The PORVs are to be considered OPERABLE whenever these MCCs are available to supply power.
- 2. Two OPERABLE RHR suction relief valves; or An RHR suction relief valve is OPERABLE for the COPS when its RHR suction isolation valve and its RHR suction valve are open, its setpoint is at or between 440 psig and 460 psig, and testing has proven its ability to open at this setpoint.
- 3. One OPERABLE PORV and one OPERABLE RHR suction relief valve; or
- b. A depressurized RCS and an RCS vent.
An RCS vent is OPERABLE when open with an area of
- 1.5 square inches (based on an equivalent length of 10 feet of pipe, i.e., capable of relieving 685 gpm at 722 psig).
Each of these methods of overpressure prevention is capable of mitigating the limiting COPS transient. APPLICABILITY This LCO is applicable in MODE 4, in MODE 5, and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits in' MODES 1, 2, and 3. When the reactor vessel head is off, overpressurization cannot occur. I (continued) Vogtle Units 1 and 2 B 3.4.12-8
COPS B 3.4.12 BASES APPLICABILITY LCO 3.4.3 provides the operational PIT limits for all MODES. (continued) LCO 3.4.10, "Pressurizer Safety Valves,' requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3. Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event. The Applicability is modified by a Note stating that accumulator isolation is only required when the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves. This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions. ACTIONS Two Notes modify the ACTIONS table. Note 1 prohibits entry into MODE 6 with the vessel head on from MODE 6 and MODE 5 from MODE 6 with the vessel head on. Entry into MODE 4 from MODE 5 is already prohibited by LCO 3.0.4. Note 2 permits entry into MODE 4 from' MODE 3 with a PORV that is inoperable for the purpose of cold overpressure protection provided that RCS temperature is maintained above 2750F, and, within 36 hours, either: the PORV is restored to OPERABLE status; or, an RHR suction relief valve is placed in service so that the requirements of LCO 3.4.12 are met. Otherwise, the reactor vessel must be depressurized and vented in accordance With Required Action F.1. With only one PORV OPERABLE, the COPS remains capable of mitigating a design basis cold overpressurization event. However, the system cannot withstand a single failure of the remaining PORV. The current COPS enable temperature is established very conservatively at 350 0F. However, the application of ASME Code Case N-514 would allow the enable temperature to be lowered to less than 2750F. Therefore, when entering this LCO from MODE 3 with one required PORV inoperable, maintaining RCS temperature above 2750 F minimizes actual exposure to a cold overpressure event. Furthermore, requiring action within 36 hours minimizes the exposure to a single failure while allowing sufficient time to either restore the inoperable PORV or to place RHR in service. Note 2 is only applicable to the condition of entering MODE 4 from MODE 3 with one required PORV inoperable for the purpose of cold overpressure protection. If operating in MODE 4 and a failure of a required RCS relief valve occurs, Condition D applies. (continued) Vogtle Units 1 and 2 B 3.4.12-9
COPS B 3.4.12 BASES ACTIONS A.1 (continued) With one or more safety injection pumps capable of injecting into the RCS, RCS overpressurization is possible. Rendering the safety injection pumps incapable of injecting into the RCS within 4 hours to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition. B.1. C.1. and C.2 An unisolated accumulator requires isolation within 1 hour. This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves. If isolation is needed and cannot be accomplished in 1 hour, Required Action C.1 and Required Action C.2 provide two options, either of which must be performed in the next 12 hours. By increasing the RCS temperature to > 350 0F, an accumulator pressure of 678 psig cannot exceed the COPS limits if the accumulators are fully injected. Depressurizing the accumulators below the COPS limit from the PTLR also gives this protection. The Completion Times are based on operating experience that these activities can be accomplished in these time periods and that the likelihood that an event requiring COPS during this time is small. D.1 In MODE 4, with one required RCS relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within a Completion Time of 7 days. Two RCS relief valves in any combination of the PORVS and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component. The Completion Time considers the facts that only one of the RCS relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. (continued) Vogtle Units 1 and 2 B 3.4.12-10
COPS B 3.4.12 BASES ACTIONS E.1 (continued) The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 7). Thus, with one of the two RCS relief valves inoperable in MODE 5 or in MODE 6 with the head on, the Completion Time to restore two valves to OPERABLE status is 24 hours. The Completion Time represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events. F.1 The RCS must be depressurized and a vent must be established within 12 hours when:
- a. Both required RCS relief valves are inoperable; or
- b. A Required Action and associated Completion Time of Condition A, C, D, or E is not met; or
- c. The COPS is inoperable for any reason other than Condition A, B, C, D, or E.
The vent must be sized 2 1.5 square inches (based on an equivalent length of 10 feet of pipe) to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel. The Completion Time considers the time required to place the unit in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements. (continued) Vogtle Units I and 2 B 3.4.12-1 1
COPS B 3.4.12 BASES SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, both safety injection pumps are verified incapable of injecting into the RCS, and the accumulator discharge isolation valves are verified closed and locked out. The safety injection pumps are rendered incapable of injecting into the RCS through at least two independent means such that a single failure or single action will not result in an injection into the RCS. The Frequency of within 4 hours after initial entry into MODE 4 from MODE 3 and prior to RCS cold leg temperature decreasing below 3250F (for the safety injection pumps) and 12 hours thereafter (for the safety injection pumps and accumulators) is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment. SR 3.4.12.3 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and by testing it in accordance with the Inservice Testing Program. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO. For Train A, the RHR suction relief valve is PSV-8708A and the suction isolation valves are HV-8701A and B. ForTrain B, the RHR suction relief valve is PSV-8708B and the suction isolation valves are HV-8702A and B. The RHR suction valves are verified to be opened every 12 hours. The Frequency is considered adequate in view of other administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction isolation valves remain open. The ASME Code, Section Xl (Ref. 8), test per Inservice Testing Program verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint. SR 3.4.12.4 The RCS vent of 2 1.5 square inches (based on an equivalent length of 10 feet of pipe) is proven OPERABLE by verifying its open condition either: (continued) Vogtle Units 1 and 2 B 3.4.12-12
COPS B 3.4.12 BASES SURVEILLANCE SR 3.4.12.4 (continued) REQUIREMENTS
- a. Once every 12 hours for a valve that cannot be locked.
- b. Once every 31 days for a valve that is locked, sealed, or secured in position. A removed pressurizer safety valve fits this category.
The passive vent arrangement must only be open to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12 b. SR 3.4.12.5 The PORV block valve must be verified open every 72 hours to provide the flow path for each required PORV to perform its function when actuated. The valve must be remotely verified open in the main control room. This Surveillance is performed if the PORV satisfies the LCO. The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation. The 72 hour Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open. SR 3.4.12.6 Performance of a COT is required within 12 hours after decreasing RCS temperature to
- 3500F and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required.
A Note has been added indicating that this SR is required to be performed 12 hours after decreasing RCS cold leg temperature to
- 3500F. The 12 hours considers the unlikelihood of a low temperature overpressure event during this time.
(continued) Vogtle Units 1 and 2 B 3.4.12-13
- -~'1.
>t:
PRESSURE TEMPERATURE LIMITS REPORT Southern Nuclear Company Vogtle Unit 1 Pressure Temperature Limits Report Revision 2, February 2004
PRESSURE TEMPERATURE LINMITS REPORT Table Of Contents List Of Tables .......... ii List Of Figures . .... i 1.0 RCS Pressure Temperature Limits Report (PTLR) .1 2.0 Operating Limits.1 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) . 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) . 3.1 Pressurizer PORV Setpoints .2 4.0 Reactor Vessel Material Surveillance Program................................................................................2 5.0 Supplemental Data Tables .3 6.0 References .19 ii
PRESSURE TEMPERATURE LIMITS REPORT List Of Tables Table 2-1 Vogtle Unit I Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .......................................................... 6 Table 2-2 Vogtle Unit I Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .7 Table 3-1 Vogtle Unit I Data Points for COPS PORV Setpoints .8 Table 5-1 Comparison of the Vogtle Unit I Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions .10 Table 5-2 Calculation of Chemistry Factors using Vogtle Unit I Surveillance Capsule Data . 11 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 1....... 12 Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit I (10 19 n/cm 2 , E > 1.0 MeV) ......................... 13 Table 5-5 Vogtle Unit I Calculation of the Adjusted Reference Temperature (ART) Values for the 1/4T Location @ 36 EFPY .14 Table 5-6 Vogtle Unit I Calculation of the ART Values for the 3/4T Location @ 36 EFPY . 15 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Vogtle Unit I I-leatup/Cooldown Curves .16 Table 5-8 RTm Calculations for Vogtle Unit I Beltline Region Materials at 36 EFPY . 17 Table 5-9 RTp5 Calculations for Vogtle Unit I Beltline Region Materials at 54 EFPY. 1 8 iii
PRESSURE TENMPERATURE LINIITS REPORT List Of Figures Figure 2-1 Vogtle Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of I00 0 F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .................................................... 4 Figure 2-2 Vogtle Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors)....................................................................5 Figure 3-1 Vogtle Unit I Maximum Allowable Nominal PORV Setpoints for COPS . 9 iv
PRESSURE TENMPERATURE LIMITS REPORT 1.0 RCS Prcssurc Temperature Limits Report (PTLR) This PTLR for Vogtle Unit I has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS) Revisions to the PTLR shall be provided to the NRC after issuance. 2.0 RCS Pressure and Temperature (P/T) Limits The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-14040, Revision 4 1'1 with exception of WCAP-16142-P, Revision 1121 (Elimination of the Flange Requirement). The operability requirements associated with COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6. 2.1 RCS P/T Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 600 F 2.1.2 The RCS temperature rate-of-change limits are:
- a. A maximum heatup rate of 100 0F in any one hour period.
- b. A maximum cooldown rate of 100F in any one hour period.
- c. A maximum temperature change of less than or equal to 10F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2. 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) The setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsections which follow. These setpoints have been developed using the NRC-approved methodology specified in TS 5.6.6. I
PRESSURE TEMPERATURE LIMNIITS REPORT 3.1 Pressurizer PORV Setpoints The pressurizer PORV setpoints are specified in Figure 3-1 and Table 3-1. The limits for the COPS setpoints are contained in the 36 EFPY steady-state curves (Table 2-2), which are beltline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 74 psi. Note: These setpoints include an allowance for the 500 F thermal transport effect for heat injection transients. A calculation has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints. 4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in UFSAR Table 5.3.1-8. The results of these examinations shall be used to update Figures 2-1, 2-2, and 3-1. The pressure vessel steel surveillance program (WCAP-l 101 l141) is in compliance with Appendix Hi3" to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23 [5. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640161 of Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G."Fracture Toughness Criteria for Protection Against Failure'" 71 . The surveillance capsule removal schedule meets the requirements of ASTM E 185-82181. The removal schedule is provided in UFSAR Table 5.3.1-8. 2
PRESSURE TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 fl-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2191, predictions. Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 1.02. Table 5-3 provides the required Vogtle Unit I reactor vessel toughness data. Table 54 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation. Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 36 EFPY for each beltline material in the Vogtle Unit I reactor vessel. The limiting beltline material was the intermediate shell plate B8805-2. Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Vogtle Unit I reactor vessel beltline materials at the I/4T and 3/4T locations for 36 EFPY. Table 5-8 provides RTPTS values for Vogtle Unit I at 36 EFPY. Table 5-9 provides RTpTs values for Vogtle Unit I at 54 EFPY 3
PRESSURE TENMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8805-2 LIMITING ART VALUES AT 36 EFPY: 1/4T, II 0lF 3/4T, 95WF 2500 2250 2000 1750 0~ in 1500 03 EL 1250 C) 1000 C., 750 500 250 0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2-1 Vogtlc Unit I Reactor Coolant System l1catup Limitations (Ileatup Rate of 100lF/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) (PlottedDataprovided on Table 2-1) 4
PRESSURE TEMPERATURE LINIMITS REPORT LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8805-2 LIMITING ART VALUES AT 36 EFPY: 1/4T, 1 0oF 3/4T, 95 0 F 2500 2250 2000 1750 C/) IL 1500 I-Cn 0 E4- 1250 la 5
,1000 g
750 500 250 0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2-2 Vogtle Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1000 F/hr) Applicable for the First 36 EFPY (WVithout Margins for Instrumentation Error) (PlottedDataprovided onl Table 2-2) 5
PRESSURE TENI PERATURE LIMITS REPORT Table 2-1 Vogtle Unit I Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) 60rF/fir ileattip 60'F/hr Ileattip 100l/hr Ileatup 100'F/hr Ileattip Leak Test Limit Criticality Limit _ Criticality Limit l T P T P T P T P T r 60 0 170 0 60 0 170 0 153 2000 60 747 170 760 60 730 170 730 170 2485 65 760 170 760 65 730 170 730 70 760 170 760 70 730 170 730 75 760 170 760 75 730 170 730 80 760 170 763 80 730 170 730 85 763 170 770 85 730 170 730 90 770 170 782 90 730 170 730 95 782 170 796 95 730 170 733 100 796 170 815 100 733 170 739 105 815 170 836 105 739 170 747 110 836 170 862 110 747 170 759 115 862 170 891 115 759 170 774 120 891 170 925 120 774 170 791 125 925 170 962 125 791 170 812 130 962 175 1005 130 812 175 837 135 1005 180 1052 135 837 180 865 140 1052 185 1105 140 865 185 897 145 1105 190 1163 145 897 190 933 150 1163 195 1228 150 933 195 974 155 1228 200 1300 155 974 200 1020 160 1300 205 1380 160 1020 205 1071 165 1380 210 1468 165 1071 210 1128 170 1468 215 1566 170 1128 215 1191 175 1566 220 1674 175 1191 220 1261 180 1674 225 1793 180 1261 225 1339 185 1793 230 1925 185 1339 230 1426 190 1925 235 2070 190 1426 235 1521 195 2070 240 2231 195 1521 240 1627 200 2231 245 2408 200 1627 245 1743 205 2408 205 1743 250 1872 210 1872 255 2014 215 2014 260 2171 220 2171 265 2344 225 2344 6
PRESSURE TEMPERATURE LIMNITS REPORT Table 2-2 Vogtle Unit I Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) Steady State 20F/hr I 40°F/hr 60F/hr ____f__r T P T P T P T I T 60 0 60 0 60 0 60 0 60 0 60 747 60 709 60 670 60 633 60 559 65 762 65 725 65 688 65 652 65 582 70 778 70 742 70 707 70 673 70 608 75 796 75 762 75 728 75 696 75 637 80 816 80 783 80 752 80 722 80 668 85 838 85 807 85 778 85 751 85 704 90 862 90 834 90 807 90 783 90 743 95 889 95 863 95 840 95 819 95 787 100 918 100 895 100 875 100 858 100 835 105 951 105 931 105 915 105 902 105 889 110 987 110 971 110 959 110 950 110 948 115 1027 115 1015 115 1007 115 1004 120 1071 120 1063 120 1061 125 1120 125 1117 130 1173 135 1233 140 1299 145 1371 150 1452 155 1541 160 1639 165 1747 170 1867 175 2000 180 2146 185 2308 7
PRESSURE TEMPERATURE LIMITS REPORT Table 3-1 Vogtle Unit 1Data Points for the Maximum Allowable Nominal COPS PORV Setpoints 8
PRESSURE TENIPERATURE LIMIITS REPORT 900 850 F-- 800 5 w cj- 750 L 700 650 600 0 550 500 450 400 0 50 100 150 200 250 300 350 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F) Figure 3-1: Vogtlc Unit I Maximum Allowable Nominal PORN' Setpoints for COPS 9
PRESSURE TEMIPERATURE LINIITS REPORT Table 5-1 Comparison of the Vogtle Unit 1 Surveillance Material 30 fl-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 1019 n/cm2) (OF) (aF) (b) (0/) (a) (%)(c) Intermediate Shell U 0.3691 27.8 13.6 15 0 Plate B8805-3'e) Y 1.276 41.0 31.9 20 0 (Longitudinal) V 2.178 46.5 42.7 23 3 Intermediate Shell U 0.3691 27.8 o(d) 15 0 Plate B8805-3(e) Y 1.276 41.0 15.2 20 0 (Transverse) V 2.178 46.5 33.8 23 2 Weld Metal t U 0.3691 25.0 25.0 15 0 Y 1.276 36.8 7.7 20 V 2.178 41.8 o(d)23 2 HAZ Metal U 0.3691 - (e0 0 (d) -- 5 Y 1.276 -() 20.8 9 V 2.178 __e 42.1 11 Notes: (a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material. (b) Calculated using measured Charpy data plotted using CVGRAPII, Version 4. Il (c) Values are based on the definition of upper shelf energy given in ASTM El185-82 18 1. (d) The actual measured value of ARTNDT for the intermediate shell plate (capsule U) is -9.58, the actual measured value of ARTNDT for the wveld metal (capsule V) is -1.34 and the actual measured value of ARTNDT for the HAZ metal (capsule U) is -19.35. This physically should not occur, therefore for conservatism a value of zero wvill be reported (i.e. No Change in T30). (e) The heat number for lower shell plate B8805-3 is C-0623-1. (n The Surveillance weld was fabricated from Wire Heat No. 83653, Flux Type Linde 0091, Flux Lot No. 3536. 10
PRESSURE TEMPERATURE LIMI TS REPORT Table 5-2 Calculation of Chemistry Factors using Vogtle Unit I Surveillance Capsule Data Material Capsule Capsule j(a) FF(b) ARTNDT(0) FF*ARTNDTr FF2 Intermediate Shell U 0.3691 0.725 13.6 9.9 0.526 Plate B8805-3 0 Y 1.276 1.068 31.9 34.1 1.141 (Longitudinal) V 2.178 1.211 42.7 51.7 1.467 Intermediate Shell U 0.3691 0.725 0(C) 0.0 0.526 Plate B8805-3() Y 1.276 1.068 15.2 16.2 1.141 (Transverse) V 2.178 1.211 33.8 40.9 1.467 SUM 152.8 6.268 CFB8805-3 = X(FF
- RTNDT) . (FF 2 ) = (152.8) + (6.268) = 24.40 F Surveillance Weld U 0.3691 0.725 25.5 (2 5 .0)(d) 18.5 0.526 Metal(g)Y 1.276 1.068 7.9 (7 . 7 )(d) 8.4 1.141 V 2.178 1.211 0(C) 0.0 1.467 SUM 26.9 3.134 CF%%Ydd = E(FF
- RTNDT)
- E(FF ) = (26.9) * (3.134) = 8.6 0 F Notes:
(a) f= Calculated fluence from capsule V dosimetry analysis results (1), (x 1019 n/cm2 , E > 1.0 MeV). (b) FF = fluence factor==0.28 -O. log fl (c) ARTNDT values are the measured 30 ft-lb shift values taken from App. C of Ref. 10, rounded to one decimal point. (d) The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 1.02. (e) Actual values for ARTNDT are -9.58 (Plate) and -1.34 (Weld). This physically should not occur, therefore for conservatism a value of zero will be used for this calculation.. (f) The heat number for lower shell plate B8805-3 is C-0623-1. (g) Surveillance Weld was fabricated from Wire Heat No.83653, Flux Type Linde 0091, Flux Lot No. 3536. l1
PRESSURE TENMPERATURE LINM ITS REPORT Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit I Material Description Ci (%) Ni(%) Initial RTNDT Closure Head Flange B8801-1 0.70 20 0 F (Heat # 123J173VAI) Vessel Flange B8802-1 0.71 OOF (Hleat # 123H402VA1) Intermediate Shell Plate B8805-1 0.083 0.597 OOF (Heat # C-0613-1) Intermediate Shell Plate B8805-2 0.083 0.61 20°1 (Heat # C-0613-2) Intermediate Shell Plate B8805-3 0.062 0.598 30OF (Heat # C-0623-1) Lower Shell Plate B8606-1 0.053 0.593 20OF (Heat # C-2146-1) Lower Shell Plate B8606-2 0.057 0.60 20WF (Heat # C-2146-2) Lower Shell Plate B8606-3 0.067 0.623 10F (Heat # C-2085-2) Intermediate Shell Longitudinal 0.042 0.102 -80 0 F Welds, 101 -124A, B & C(b) Lower Shell Longitudinal Welds, 0.042 0.102 -80 0 F 101-142A, B & C(bl Circumferential Weld 1 01-1 7 1 (b) 0.042 0.102 -80 0 F Surveillance Program Weld Metal(b) 0.040 0.102 Notes: (a) The initial RTNDT values for the plates and welds are based on measured data. (b) All welds, including the surveillance weld, were fabricated with weld wfire heat number 83653, Linde 0091 Flux, Lot No. 3536. Per Regulatory Guide 1.99, Revision 2, "weight percent copper " and "weight percent nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld." 12
PRESSURE TENIPERATURE LIMITS REPORT Table 5-4 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit I (I0O" n/cm 2 , E > 1.0 MeV) Azimuthal Location EFPY 00 150 12.50 NP 200 NP 22.50 NP 450 300 300 300 8.57 0.290 0.433 0.533 0.349 0.289 0.523 16 0.527 0.781 0.953 0.624 0.517 0.936 36 1.17 1.72 2.09 1.36 1.13 2.05 54 1.74 2.56 3.10 2.03 1.68 3.05 13
PRESSURE TENI PERATURE LINI ITS REPORT Table 5-5 Vogtle Unit I Calculation of the ART Values for the l/4T Location @ 36 EFPYV0 Material RG 1.99 R2 CF IFF IRTNDT('2 ARTNDr(b) Alargin"' ART(d) Method ( 0F) Intermediate Shell Plate B8805-1 Position 1.1 53.1 1.06 0 56.3 34 90 Intermediate Shell Plate B8805-2 Position 1.1 53.1 1.06 20 56.3 34 110 Intermediate Shell Plate B8805-3 Position 1.1 38.4 1.06 30 40.7 34 105 Position 2.1 24.4 1.06 30 26.0 17(c) 73 Lower Shell Plate B8606-1 Position 1.1 32.8 1.06 20 34.8 34 89 Lower Shell Plate B8606-2 Position 1.1 35.2 1.06 20 37.3 34 91 Lower Shell Plate B8606-3 Position 1.1 41.9 1.06 10 44.4 34 88 Inter. Shell Longitudinal Weld Position 1.1 34.5 0.899 -80 31.0 -18 Seam 101-124A (0° Azimuth) Position 2.1 8.6 0.899 -80 7.7 7.7(e) -65 Inter. Shell Long. Weld Seams Position 1.1 34.5 1.06 -80 36.6 36.6 -7 101-124B,C Position 2.1 8.6 1.06 -80 9.1 9.1 (e) -62 (1200, 2400Azinjuth) Intermediate to Lower Shell Position 1.1 34.5 1.06 -80 36.6 36.6 -7 Girth Weld Seam 101 -171 Position 2.1 8.6 1.06 -80 9.1 9.1 (e) -62 Lower Shell Long. Weld Seams Position 1.1 34.5 1.06 -80 36.6 36.6 -7 101 - 142A,C (600. 3000 Azimuth) Position 2.1 8.6 1.06 -80 9.1 9. (e) -62 Lower Shell Long. Weld Seam Position 1.1 34.5 0.899 -80 31.0 31 -18 101-142B (180° Azimuith) Position 2.1 8.6 0.899 -80 7.7 7.7(c) -65 Notes: (a) Initial RTNDT values are measured values. (b) ARTNDT = CF
- FF (c) M = 2 *(cj2+ FA2)12 (d) ART = Initial RTNDT + ARTNDT + Margin (°F); (Rounded per ASTM E29, using the "Rounding Method").
(e) Data deemed credible per Reference 10. (f Neutron Fluence value used for all material is the highest value (@ 300) from Table 5-4 for 36 EFPY with exception to intermediate shell longitudinal weld 101-124A and lower shell longitudinal weld 10I1-142B which used the fluence at 00 from Table 5-4 for 36 EFPY. 14
PRESSURE TENMPERATURE LINIITS REPORT Table 5-6 Vogtle Unit I Calculation of the ART Values for the 3/4T Location @ 36 EFPY t0 Material RG 1.99 R2 CF FF IRTNDT'2' ARTNDT() largin<'c ART/d) Method (0 F) Intermediate Shell Plate B8805-1 Position 1.1 53.1 0.773 0 41.0 34 75 Intermediate Shell Plate B8805-2 Position 1.1 53.1 0.773 20 41.0 34 95 Intermediate Shell Plate B8805-3 Position 1.1 38.4 0.773 30 29.7 29.7 89 Position 2.1 24.4 0.773 30 18.9 17(c) 66 Lower Shell Plate B8606-1 Position 1.1 32.8 0.773 20 25.4 25.4 71 Lower Shell Plate B8606-2 Position 1.1 35.2 0.773 20 27.2 27.2 74 Lower Shell Plate B8606-3 Position 1.1 41.9 0.773 10 32.4 32.4 75 Inter. Shell Longitudinal Weld Position 1.1 34.5 0.622 -80 21.5 21.5 -37 Seam 101 - 124A (0° Azimuth) Position 2.1 8.6 0.622 -80 5.3 53(C) -69 Inter. Shell Long. Weld Seams Position 1.1 34.5 0.773 -80 26.7 26.7 -27 101-124B,C Position 2.1 8.6 0.773 -80 6.6 6.6(c) -67 (120°, 2400 Azimutth) Intermediate to Lower Shell Position 1.1 34.5 0.773 -80 26.7 26.7 -27 Girth Weld Seam 101-171 Position 2.1 8.6 0.773 -80 6.6 6.6(c) -67 Lower Shell Long. Weld Seams Position 1.1 34.5 0.773 -80 26.7 26.7 -27 101-142A,C (60°, 3000 Azimuth) Position 2.1 8.6 0.773 -80 6.6 6.6(e) -67 Lower Shell Long. WVeld Seam Position 1.1 34.5 0.622 -80 21.5 21.5 -37 101-142B (180° Azimuth) Position 2.1 8.6 0.622 -80 5.3 5.3(c) -69 Notes: (a) Initial RTNDT values are measured values. (b) ARTNDT = CF
- FF (c) M = 2 *(cj2 + 2s,,2 (d) ART = Initial RTNDT + ARTNDT + Margin (°F); (Rounded per ASTM E29, using the "Rounding Method").
(e) Data deemed credible per Reference 10. (f) Neutron Fluence value used for all material is the highest value (@a 30°) from Table 5-4 for 36 EFPY with exception to intermediate shell longitudinal weld 101-124A and lower shell longitudinal weld 101-142B which used the fluence at 0° from Table 54 for 36 EFPY. 15
PRESSURE TENIPERATURE LINM ITS REPORT Table 5-7 Summary of the Vogtle Unit I Reactor Vessel Beltline Material ART Values Material 1RG 1.99 R2 l %ART 3/4 ART Method (OF) (0F) Intermediate Shell Plate B8805-1 Position 1.1 90 75 Intermediate Shell Plate B8805-2 Position 1.1 110 95 Intermediate Shell Plate B8805-3 Position 1.1 105 89 Position 2.1 73 66 Lower Shell Plate B8606-1 Position 1.1 89 71 Lower Shell Plate B8606-2 Position 1.1 91 74 Lower Shell Plate B8606-3 Position 1.1 88 75 Inter. Shell Longitudinal Weld Position 1.1 -18 -37 Seam 101-124A (00 Azimuth) Position 2.1 -65 -69 Inter. Shell Long. Weld Seams Position 1.1 -7 -27 101-124B,C Position 2.1 -62 -67 (1200, 2400Azintnth) Intermediate to Lower Shell Position 1.1 -7 -27 Girth Weld Seam 10 -171 Position 2.1 -62 -67 Lower Shell Long. Weld Seams Position 1.1 -7 -27 101-142A,C (60°, 3000 Azinnith) Position 2.1 -62 -67 Lower Shell Long. Weld Seam Position 1.1 -18 -37 101-142B (1800 Azinnith) Position 2.1 -65 -69 16
PRESSURE TENI PERATURE LIMIITS REPORT Table 5-8 RTp~ys Calculations for Vogtle Unit I Beltline Region Materials at 36 EFPYV Material RG 1.99 R2 CF FF IRTNDT(UT) AR l arginS RTii s(d) MNtethod (OF) Intermediate Shell Plate B8805-1 Position 1.1 53.1 1.20 0 63.7 34 98 Intermediate Shell Plate B8805-2 Position 1.1 53.1 1.20 20 63.7 34 118 Intermediate Shell Plate B8805-3 Position 1.1 38.4 1.20 30 46.1 34 110 Position 2.1 24.4 1.20 30 29.4 17(') 76 Lower Shell Plate B8606-1 Position 1.1 32.8 1.20 20 39.4 34 93 Lower Shell Plate B8606-2 Position 1.1 35.2 1.20 20 42.2 34 96 Lower Shell Plate B8606-3 Position 1.1 41.9 1.20 10 50.3 34 94 Inter. Shell Longitudinal Weld Position 1.1 34.5 1.04 -80 35.9 35.9 -8 Seam 101-124A (00 Azimnuth) Position 2.1 8.6 1.04 -80 8.9 8.9(e) -62 Inter. Shell Long. Weld Seams Position 1.1 34.5 1.20 -80 41.4 41.4 3 101-124B, C Position 2.1 8.6 1.20 -80 10.3 10.3(c) 59 (1200, 2400Azimnuth) . Intermediate to Lower Shell Position 1.1 34.5 1.20 -80 41.4 41.4 3 Girth Weld Seam 101-171 Position 2.1 8.6 1.20 -80 10.3 10.3(e) -59 Lower Shell Long. Weld Seams Position 1.1 34.5 1.20 -80 41.4 41.4 3 101-142A, C (600, 3000 Azimuth) Position 2.1 8.6 1.20 -80 10.3 10.3(') -59 Lower Shell Long. Weld Seam Position 1.1 34.5 1.04 -80 35.9 35.9 -8 101-142B (1800 Azimtah) Position 2.1 8.6 1.04 -80 8.9 8.9(c) -62 Notes: (a) Initial RTNDT values are measured values (b) ARTPTS = CF
- FF (c) M = 2 *(oi 2 + CA2)1/2 (d) RTPTS = RTNDT(LJ) + ARTrrs + Margin (fF)
(e) Data deemed credible per Reference 10. (I) Neutron Fluence value used for all material is the highest value (@ 30°) from Table 5-4 for 36 EFPY with exception to intermediate shell longitudinal weld 101-124A and lower shell longitudinal weld 10 1-142B which used the fluence at 00 from Table 5-4 for 36 EFPY. 17
PRESSURE TEMPERATURE LINITS REPORT Table 5-9 RTpTs Calculations for Vogtle Unit I Beltline Region Materials at 54 EFPY"') Material RG 1.99 R2 CF FF IRTNDT(t)) ARTrrs() Margin(c) RTrrsfd) Method (0 F) Intermediate Shell Plate B8805-1 Position 1.1 53.1 1.30 0 69.0 34 103 Intermediate Shell Plate B8805-2 Position 1.1 53.1 1.30 20 69.0 34 123 Intermediate Shell Plate B8805-3 Position 1.1 38.4 1.30 30 49.9 34 114 Position 2.1 24.4 1.30 30 31.7 17(e) 79 Lower Shell Plate B8606-1 Position 1.1 32.8 1.30 20 42.6 34 97 Lower Shell Plate B8606-2 Position 1.1 35.2 1.30 20 45.8 34 100 Lower Shell Plate B8606-3 Position 1.1 41.9 1.30 10 54.5 34 99 Inter. Shell Longitudinal Weld Position 1.1 34.5 1.15 -80 39.7 39.7 0 Seam 101-124A (0° Azinnmth) Position 2.1 8.6 1.15 -80 9.9 9.9(°) -60 Inter. Shell Long. Weld Seams Position 1.1 34.5 1.30 -80 44.9 44.9 10 101-124B, C Position 2.1 8.6 1.30 -80 11.2 11.2('e) -58 (1200° 2400Azimuth) Intermediate to Lower Shell Position 1.1 34.5 1.30 -80 44.9 44.9 10 Girth Weld Seam 101-171 Position 2.1 8.6 1.30 -80 11.2 11.2('e) -58 Lower Shell Long. Weld Seams Position 1.1 34.5 1.30 -80 44.9 44.9 10 101-142A, C (60°, 3000 Azimuth) Position 2.1 8.6 1.30 -80 11.2 11.2(e) -58 Lower Shell Long. Weld Seam Position 1.1 34.5 1.15 -80 39.7 39.7 0 101-142B (180° Azimtah) Position 2.1 8.6 1.15 -80 9.9 9.9( -60 Notes: (a) Initial RTNDT values are measured values (b) ARTPTS = CF
- FF (c) M =2 (i2 + c 2)12 (d) RTP'rs = RTNDT(U) + ARTr~s + Margin (0F)
(e) Data deemed credible per Reference 10. (f Neutron Fluence value used for all material is the highest value (@ 300) from Table 5-4 for 36 EFPY with exception to intermediate shell longitudinal weld 101-124A and lower shell longitudinal weld 101- 142B which used the fluence at 0° from Table 5-4 for 36 EFPY. 18
PRESSURE TEMPERATURE LIMITS REPORT 6.0 References
- 1. WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al.
- 2. WCAP-16142-P, Revision 1,"Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units I and 2", Warren Bamford, et. al., February 2004.
- 3. Code of Federal Regulations, I0CFR50, Appendix 1H, Reactor Vessel AIaterial Surveillance ProgramRequirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 4. WCAP- I1011, GeorgiaPower Coinpanj'Alvin IT.' Vogde Unit No. 2 Reactor Vessel Radiation Surveillance Program,L. R. Singer, February 1986.
- 5. ASTM E23 Standard Test Alethod Notched Bar nmpact Testing of Metallic AMaterials, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 6. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
- 7. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G. FractlureToughness CriteriaforProtectionAgainst Failure.
- 8. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, StandardPractice for ConductingSurveillance Testsfor Light- Water Cooled Nuclear PowerReactor 1essels.
- 9. Regulatory Guide 1.99, Revision 2, Radiation Embrittlemnent of Reactor Vessel Materials,U.S.
Nuclear Regulatory Commission, May 1988.
- 10. WCAP-1 5067, Anal'sis of Capsile V From the Southern Nuclear lVogtle Electric Generating Plant Unit I Reactor Vessel Radiation Surveillance Program, T.J. Laubliam, et. al., Dated September 1998. [Note that the TestingMnalysis reportsforsurveillance capsiules U and Yfrom Vlogtle Unit I were documented tunder JWCAP-12256 and 111CAP-13931, Rei: 1, respectively<]
- 11. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.
19
PRESSURE TENMPERATURE LIM ITS REPORT Southern Nuclear Company Vogtle Unit 2 Pressure Temperature Limits Report Revision 2, February 2004
PRESSURE TEMPERATURE LIMITS REPORT Table Of Contents List Of Tables ..... 111ii List Of Figures ........... iv 1.0 RCS Pressure Temperature Limits Report (PTLR) .I 2.0 Operating Limits.1 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) .1 3.0 Cold Overpressure Protection Systems (COPS) (LCO 3.4.12) . 3.1 Pressurizer PORV Setpoints .2 4.0 Reactor Vessel Material Surveillance Program .2 5.0 Supplemental Data Tables .3 6.0 References .19 ii
PRESSURE TEMPERATURE LIMIITS REPORT List Of Tables Table 2-1 Vogtle Unit 2 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .......................................................... 6 Table 2-2 Vogtle Unit 2 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .7 Table 3-1 Vogtle Unit 2 Data Points for COPS PORV Setpoints .8 Table 5-1 Comparison of the Vogtle Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions .10 Table 5-2 Calculation of Chemistry Factors using Vogtle Unit 2 Surveillance Capsule Data .1 1 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 2 ....... 12 Table 54 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 2 (10"9 n/cm2 , E > 1.0 MeV) .............. 13 Table 5-5 Vogtle Unit 2 Calculation of the Adjusted Reference Temperature (ART) Values for the l/4T Location @ 36 EPY.14 Table 5-6 Vogtle Unit 2 Calculation of the ART Values for the 3/4T Location @ 36 EFPY . 15 Table 5-7 Summary of the Vogtle Unit 2 Reactor Vessel Beltline Material ART Values 16 Table 5-8 RTPTS Calculations for Vogtle Unit 2 Beltline Region Materials at 36 EFPY . 17 Table 5-9 RTpTs Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY . 18 iii
PRESSURE TENMPERATURE LIM ITS REPORT List Of Figures Figure 2-1 Vogtle Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100'F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) ................................ : . 4 Figure 2-2 Vogtle Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to I000 F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .................................................... 5 Figure 3-1 Vogtle Unit 2 Maximum Allowable Nominal PORV Setpoints for COPS ......................... 9 iv
PRESSURE TENIPERATURE LIMI ITS REPORT 1.0 RCS Pressure Temperature Limits Report (PTLR) This PTLR for Vogtle Unit 2 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS) Revisions to the PTLR shall be provided to the NRC after issuance. 2.0 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-14040, Revision 4111 with exception of WCAP-16142-P, Revision 11 (elimination of the flange requirement). The operability requirements associated with the COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6. 2.1 RCS P/T Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60'F 2.1.2 The RCS temperature rate-of-change limits are:
- a. A maximum heatup rate of 1000 F in any one hour period.
- b. A maximum cooldown rate of 1000 F in any one hour period.
- c. A maximum temperature change of less than or equal to 100 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2. 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) The setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsections which follow. These setpoints have been developed using the NRC-approved methodology specified in TS 5.6.6. I
PRESSURE TEN]PERATURE LIMI ITS REPORT 3.1 Pressurizer PORV Setpoints The pressurizer PORV setpoints are specified in Figure 3-1 and Table 3-1. The limits for the COPS setpoints are contained in the 36 EFPY steady-state curves (Table 2-2), which are beltline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 74 psi. Note: These setpoints include an allowance for the 50°1; thermal transport effect for heat injection transients. A calculation has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints. 4.0 Reactor 'Vesscl M1atcrial Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in UFSAR Table 5.3.1-9. The results of these examinations shall be used to update Figures 2-1, 2-2, and 3-1. The pressure vessel steel surveillance program (WCAP-I 1381 I') is in compliance with Appendix 1I1I-to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23 15. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640161 of Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G "Fracture Toughness Criteria for Protection Against Failure"171. The surveillance capsule removal schedule meets the requirements of ASTM El 85-82181. The removal schedule is provided in UFSAR Table 5.3.1-9. 2
PRESSURE TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 fl-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2191, predictions. Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 1.19. Table 5-3 provides the required Vogtle Unit 2 reactor vessel toughness data. Table 54 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation. Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 36 EFPY for each beltline material in the Vogtle Unit 2 reactor vessel. The limiting beltline material was the lower shell plate R8-1. Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Vogtle Unit 2 reactor vessel beltline materials at the 1/4T and 3/4T locations for 36 EFPY. Table 5-8 provides RTpTs values for Vogtle Unit 2 at 36 EFPY. Table 5-9 provides RTpTs values for Vogtle Unit 2 at 54 EFPY 3
PRESSURE TEMPERATURE LINI ITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R8-1 LIMITING ART VALUES AT 36 EFPY: 1/4T, 120 0 F 3/4T, 107°1 2500 2250 2000 1750 a-1500 2 U) 41, 1250 (0 U 1000 (3 750 500 250 a 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2-1 Vogtle Unit 2 Reactor Coolant System lleatup Limitations (Hleatup Rate of 100OF/l1r) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) (PlottedDataprovided on Table 2-1) 4
PRESSURE TEMPERATURE LINM ITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R8-1 LIMITING ART VALUES AT 36 EFPY: 1/4T, 120 0 F 3/4T, 107 0 F 2500 2250 2000 1750 CD 1500 0 0 1250 a-1.6. 0 1000 C. C. 750 500 250 0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2-2 V'ogtle Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1001F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Error) (PlottedDataprovided on Table 2-2) 5
PRESSURE TEMPERATURE LIM ITS REPORT Table 2-1 Vogtle Unit 2 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) 60'F/hr lleatup 60'F/hr lleatiip 100'F/hr Ileatup 100'r/lir Ileatup Leak Test Limit Criticality Limit Criticality Limit T P T P T P T P T P 60 0 180 0 60 0 180 0 163 2000 60 717 180 734 60 684 180 684 180 2485 65 717 180 727 65 684 180 684 70 717 180 720 70 684 180 684 75 717 180 717 75 684 180 684 80 717 180 717 80 684 180 684 85 717 180 721 85 684 180 684 90 721 180 729 90 684 180 684 95 729 180 739 95 684 180 684 100 739 180 753 100 684 180 686 105 753 180 769 105 686 180 691 110 769 180 788 110 691 180 699 115 788 180 811 115 699 180 709 120 811 180 837 120 709 180 722 125 837 180 866 125 722 180 737 130 866 180 899 130 737 180 756 135 899 180 936 135 756 180 777 140 936 185 977 140 777 185 801 145 977 190 1023 145 801 190 829 150 1023 195 1074 150 829 195 860 155 1074 200 1130 155 860 200 896 160 1130 205 1193 160 896 205 935 165 1193 210 1262 165 935 210 979 170 1262 215 1339 170 979 215 1029 175 1339 220 1424 175 1029 220 1084 180 1424 225 1517 180 1084 225 1144 185 1517 230 1621 185 1144 230 1212 190 1621 235 1735 190 1212 235 1287 195 1735 240 1861 195 1287 240 1369 200 1861 245 2000 200 1369 245 1461 205 2000 250 2154 205 1461 250 1562 210 2154 255 2324 210 1562 255 1673 215 2324 215 1673 260 1796 220 1796 265 1932 225 1932 270 2082 230 2082 275 2248 235 2248 280 2430 240 2430 6
PRESSURE TEMPERATURE LINIITS REPORT Table 2-2 Vogtle Unit 2 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) Steady State 20'F/Iir 40hF/hr 60'F/hr _ 00__/hr T P I T P T P T I T p 60 0 60 0 60 0 60 0 60 0 60 722 60 681 60 640 60 599 60 518 65 734 65 694 65 654 65 614 65 537 70 747 70 708 70 670 70 632 70 557 75 762 75 724 75 687 75 651 75 581 80 778 80 742 80 706 80 672 80 606 85 796 85 761 85 728 85 695 85 635 90 816 90 783 90 752 90 721 90 667 95 838 95 807 95 778 95 750 95 703 100 862 100 833 100 807 100 782 100 742 105 889 105 863 105 839 105 818 105 786 110 918 110 895 110 875 110 858 110 834 115 951 115 931 115 914 115 901 115 888 120 987 120 971 120 958 120 950 120 948 125 1027 125 1015 125 1007 125 1003 130 1071 130 1063 130 1060 135 1120 135 1117 140 1173 145 1233 150 1299 155 1371 160 1452 165 1541 170 1639 175 1747 180 1867 185 2000 190 2146 195 195 2308 2308 7
PRESSURE TEMlPERATURE LIMITS REPORT Table 3-1 Vogtle Unit 2 Data Points for the Maximum Allowable Nominal COPS PORV Setpoints Temperature PORV Setpoint (psig) (Dcg.F) 70 580 90 580 140 612 201 760 202 760 350 760 8
PRESSURE TEMPERATURE LIMITS REPORT 900 850 0 800 cn 750 700 0 650 600 0
- z. 550 500 450 400 0 50 100 150 200 250 300 350 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F)
Figurc 3-1: \'ogtIc Unit 2 Maximum Allowable Nominal POR' Setpoints for COPS 9
PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 Comparison of the Vogtle Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 10i9 1/cm2) (OF) (a) (OF) (b) (%) m(O) Lower Shell<') U 0.397 23.06 2.12 15 0 Plate B8628-1 Y 1.27 33.17 5.76 20 0 (Longitudinal) X 2.01 36.89 29.35 22 3 Lower Shell(') U 0.397 23.06 0 . 0 0(d) 15 0 Plate B8628-1 Y 1.27 33.17 1.93 20 0 (Transverse) X 2.01 36.89 29.72 22 7 Weld Metaltn U 0.397 27.08 o.'d' 15 0 Y 1.27 38.95 18.59 20 7 X 2.01 43.32 20.07 22 5 HAZ Metal U 0.397 --. d)-- 0 Y 1.27 - 0 00 (d) --- 0 X 2.01 --- 0.0 0 (d) --- 7 Notes: (a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material. (b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1It '1 (c) Values are based on the definition of upper shelf energy given in ASTM E 185-82 81 . (d) Actual values for ARTNDT are -7.14 (Plate), -17.49 (Weld), -24.05 (HAZ Cap. U), -9.86 (HAZ Cap. Y) and
-2.1 (HAZ Cap. X). This physically should not occur, therefore for conservatism a value of zero will be reported (i.e. No Change in T3 0).
(e) The heat number for lower shell plate B8628-1 is C-3500-2. (f) The Surveillance weld was fabricated from Wire Heat No. 87005, Flux Type Linde 124, Flux Lot No. 1061. 10
PRESSURE TEMPERATURE LIM ITS REPORT Table 5-2 Calculation of Chemistry Factors using Vogtle Unit 2 Surveillance Capsule Data Material Capsule Capsule fz l FVt1 l ARTNDT( l FF*ARTNDT FF2 Lower Shell U 0.397 0.744 2.1 1.6 0.554 Plate B8628-1( 0 Y 1.27 1.07 5.8 6.2 1.14 (Longitudinal) X 2.01 1.19 29.4 35.0 1.42 Lower Shell U 0.397 0.744 0.0(e) 0 0.554 Plate B8628-1t f Y 1.27 1.07 1.9 2.0 1.14 (Transverse) X 2.01 1.19 29.7 35.3 1.42 SUM: 80.1 6.228 CFBS628 -1 = X (FF
- RTNDT) . ( FF 2 ) = (80.1) + (6.228) = 12.9°1 Surveillance WVeld U 0.397 0.744 0.0(e) 0 0.554 Material()Y 1.27 1.07 22.1(18.6) (d) 23.6 1.14 X 2.01 1.19 23.9(20.1)"d 28.4 1.42 SUM: 52.0 3.114 CFsurv Ovld = E(FF
- RTNDT) *. ( FF2 ) = (52.0) . (3.114) = 16.70F Notes:
(a) f = Calculated fluence from capsule X dosimetry analysis results (10), (x 1019 n/cm2, E > 1.0 MeV). (b) FF = fluence factor= fr028-0I *lo (c) ARTNDT values are the measured 30 ft-lb shift values taken from App. B of Ref. 10, rounded to one decimal point. (d) The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 1.19. (e) Actual values for ARTNDT are -7.14 (Plate) and -17.49 (Weld). This physically should not occur; therefore for conservatism a value of zero will be used. (f) The heat number for lower shell plate B8628-1 is C-3500-2. (g) Surveillance Weld was fabricated from Wire Heat No. 87005, Flux Type Linde 124, Flux Lot No. 1061. 11
PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 2 Material Description Cu (%) Ni(%) Initial RT1NDT(a) Closure Head Flange R7-1 --- 0.72 10F (Heat # 125L630VAJ) Vessel Flange Rl-l --- 0.87 -60WF Intermediate Shell Plate R4-1 0.07 0.63 10F (Heat # C-3527-1) Intermediate Shell Plate R4-2 0.06 0.61 10F (Heat # C-3527-2) Intermediate Shell Plate R4-3 0.05 0.60 30OF (Heat # C-3552-1) Lower Shell Plate B8825-1 0.06 0.62 40OF (Heat # C-3500-1) Lowver Shell Plate R8-1 0.07 0.63 400 F (Heat # C-4304-1) Lower Shell Plate B8628-1 0.05 0.59 50OF (Heat # C-3500-2) Intermediate Shell Longitudinal Weld 0.05 0.15 -100 F Seams 101-124A, B & C Lower Shell Longitudinal Weld Seams 0.05 0.15 -10 0F 101-142A, B & C Intermediate to Lower Shell Plate 0.05 0.15 -30 0 F Circumferential Weld Seam 101-171 Surveillance Weld(b) 0.04 0.13 Notes: (a) The initial RTNDT values for the plates and welds are based on measured data. (b) The weld material in the Vogtle Unit 2 surveillance program was made of the same wire and flux as the reactor vessel intermediate to lower shell girth seam weld (101-171). These welds were fabricated using weld wire heat no. 87005, Linde 124 Flux, lot no. 1061. The intermediate shell longitudinal weld seams (101-124A,B,C) and the lower shell longitudinal weld seams (101-142A,B,C) were fabricated using weld wire heat no. 87005, Linde 0091 Flux, lot no. 0145. Hence the surveillance weld is representative of all beltline welds. 12
PRESSURE TEMPERATURE LENI ITS REPORT Table 54 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 2 (I109 n/cm2 , E > 1.0 MeV) Azimuthal Location EFPY 00 150 300 450 7.62 0.263 0.381 0.456 0.452 16 0.525 0.763 0.914 0.908 36 1.15 1.67 2.01 2.01 54 1.71 2.49 2.99 2.99 13
PRESSURE TEMPERATURE LIMIITS REPORT Table 5-5 Vogtle Unit 2 Calculation of the ART Values for the 1/4T Location ( 36 EFPY( 0 Material RG 1.99 R2 CF FF IRTNIT(a) ARTNDT(b) Margin(c) ART4d) Method (OF) Intermediate Shell Plate R4-1 Position 1.1 44.0 1.051 10 46.2 34 90 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.051 10 38.9 34 83 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.051 30 32.6 32.6 95 Lower Shell Plate B8825-1 Position 1.1 37.0 1.051 40 38.9 34 113 Lower Shell Plate R8-1 Position 1.1 44.0 1.051 40 46.2 34 120 Lower Shell Plate B8628-1 Position 1.1 31.0 1.051 50 32.6 32.6 115 Position 2.1 12.9 1.051 50 13.6 13.6(c) 77 Intermediate Shell Longitudinal Position 1.1 43.3 1.051 -10 45.5 45.5 81 Weld Seams 101 -124A, B, C Position 2.1 16.7 1.051 -10 17.6 17.6(c) 25 Lower Shell Longitudinal Position 1.1 43.3 1.051 -10 45.5 45.5 81 Weld Seams 101-142A, B, C Position 2.1 16.7 1.051 -10 17.6 17.6(') 25 Intermediate to Lower Shell Position 1.1 43.3 1.051 -30 45.5 45.5 61 Circ. Weld Seam 101 -171 Position 2.1 16.7 1.051 -30 17.6 17.6(e) 5 Notes: (a) Initial RTNDT values are measured values. (b) ARTNDT = CF
- FF (c) M = 2 *(oj2 + a%2 )1"2 (d) ART = Initial RTNDT + ARTNDT + Margin (0F); (Rounded per ASTM E29, using the "Rounding Method").
(e) Data deemed credible per Reference 10. (M) Neutron Fluence value used for all material is the highest value from Table 5-4 for 36 EFPY. 14
PRESSURE TEN] PERATURE LIM ITS REPORT Table 5-6 Vogtle Unit 2 Calculation of the ART Values for the 3/4T Location @ 36 EFPY t0 Material RG 1.99 R2 CF FF IRTNIDT.2' ARTNDT(bl Nargin(~' ART(d) Method (OF) Intermediate Shell Plate R4-1 Position 1.1 44.0 0.763 10 33.6 33.6 77 Intermediate Shell Plate R4-2 Position 1.1 37.0 0.763 10 28.2 28.2 66 Intermediate Shell Plate R4-3 Position 1.1 31.0 0.763 30 23.7 23.7 77 Lower Shell Plate B8825-1 Position 1.1 37.0 0.763 40 28.2 28.2 96 Lower Shell Plate R8-1 Position 1.1 44.0 0.763 40 33.6 33.6 107 Lower Shell Plate B8628-1 Position 1.1 31.0 0.763 50 23.7 23.7 97 Position 2.1 12.9 0.763 50 9.8 9.8(c) 70 Intermediate Shell Longitudinal Position 1.1 43.3 0.763 -10 33.0 33.0 56 WVeld Seams 101-124A, B, C Position 2.1 16.7 0.763 -10 12.7 12.7(") 15 Lower Shell Longitudinal Position 1.1 43.3 0.763 -10 33.0 33.0 56 WVeld Seams 101-142A, B, C Position 2.1 16.7 0.763 -10 12.7 12.7(') 15 Intermediate to Lower Shell Position 1.1 43.3 0.763 -30 33.0 33.0 36 Circ. Weld Seam 101-171 Position 2.1 16.7 0.763 -30 12.7 12.7('c) 5 Notes: (a) Initial RTNDT values are measured values. (b) ARTNDT = CF
- FF (c) M = 2 *(o,2 + CA2)12 (d) ART = Initial RTNDT + ARTNDT + Margin (0F); (Rounded per ASTM E29, using the "Rounding Method").
(e) Data deemed credible per Reference 10. (f) Neutron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY. 15
PRESSURE TENI PERATURE LINI ITS REPORT Table 5-7 Summary of the Vogtle Unit 2 Reactor Vessel Beltline Material ART Values Material RG 1.99 I R2 Method l 1 1/4 ART (OF) I 3/4 ART (OF) Intermediate Shell Plate R4-1 Position 1.1 90 77 Intermediate Shell Plate R4-2 Position 1.1 83 66 Intermediate Shell Plate R4-3 Position 1.1 95 77 Lower Shell Plate B8825- I Position 1.1 113 96 Lower Shell Plate R8-1 Position 1.1 120 107 Lower Shell Plate B8628-1 Position 1.1 115 97 Position 2.1 77 70 Intermediate Shell Longitudinal Position 1.1 SI 56 Weld Seams 101-124A, B, C Position 2.1 25 15 Lower Shell Longitudinal Position 1.1 81 56 Weld Seams 101-142A, B, C Position 2.1 25 15 Intermediate to Lower Shell Position 1.1 61 36 Circ. Weld Seam 101-171 Position 2.1 5 -5 16
PRESSURE TEMPERATURE LIMITS REPORT Table 5-8 RTPTS Calculations for Vogtle Unit 2 Beltline Region Materials at 36 EFPYt0 Material RG 1.99 R2 CF FF IRTNDT(U)( ARTPrS(b) Margin c) RTr5 s Method (OF) Intermediate Shell Plate R4-1 Position 1.1 44.0 1.19 10 52.4 34 96 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.19 10 44.0 34 88 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.19 30 36.9 34 101 Lower Shell Plate B8825-1 Position 1.1 37.0 1.19 40 44.0 34 118 Lower Shell Plate R8-1 Position 1.1 44.0 1.19 40 52.4 34 126 Lower Shell Plate B8628-1 Position 1.1 31.0 1.19 50 36.9 34 121 Position 2.1 12.9 1.19 50 15.4 15.4'e) 81 [ oito II 4 3 104 -045.0 45.0 80 Inter. Shell Longitudinal Weld Position 1.1 43.3 1.04 -10 Seam 101-124A Position 2.1 16.7 1.04 -10 17.4 17.4(') 25 (00 Azimuth) Inter. Shell Long. Weld Seams Position 1.1 43.3 1.19 -10 51.5 51.5 93 101 -124B, C Position 2.1 16.7 1.19 -10 19.9 19.9(e) 30 (1200, 240° Azimuth) Intermediate to Lower Shell Position 1.1 43.3 1.19 -30 51.5 51.5 73 Girth Weld Seam 101-171 Position 2.1 16.7 1.19 -30 19.9 19.9(c) 10 Lower Shell Long. Weld Seams Position 1.1 43.3 1.19 -10 51.5 51.5 93 101-142B, C (210°, 330° Position 2.1 16.7 1.19 -10 19.9 19.9(C) 30 Azimuth) Lower Shell Long. Weld Seam Position 1.1 43.3 1.04 -10 45.0 45.0 80 101-142A (900 Azimuth) Position 2.1 16.7 1.04 -10 17.4 17.4(e) 25 Notes: (a) Initial RTNDT values are measured values (b) ARTPTS = CF
- FF 2
(c) M = 2*(i + a 2)1.2 (d) RTpTs = RTNDT(LI) + ARTPTS + Margin ( 0F) (e) Data deemed credible per Reference 10. (M Neutron Fluence value used for all material is the highest value from Table 5-4 for 36 EFPY. 17
PRESSURE TENI PERATURE Li Ni ITS REPORT Table 5-9 RTpTs Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY(0 Material RG 1.99 R2 CF FF IRTNDT(U) ARTPS(b Margin(') RTrrs(d) Method (OF) Intermediate Shell Plate R4-1 Position 1.1 44.0 1.29 10 56.8 34 101 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.29 10 47.7 34 92 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.29 30 40.0 34 104 Lower Shell Plate B8825-1 Position 1.1 37.0 1.29 40 47.7 34 122 Lower Shell Plate R8-1 Position 1.1 44.0 1.29 40 56.8 34 131 Lower Shell Plate B8628- 1 Position 1.1 31.0 1.29 50 40.0 34 124 Position 2.1 12.9 1.29 50 16.6 16.6(') 83 Inter. Shell Longitudinal Weld Position 1.1 43.3 1.15 -10 49.8 49.8 90 Seam 101-124A Position 2.1 16.7 1.15 -10 19.2 19.2(c) 28 (00 Azimuth) Inter. Shell Long. Weld Seams Position 1.1 43.3 1.29 -10 55.9 55.9 102 101-124B, C Position 2.1 16.7 1.29 -10 21.5 21.5'c) 33 (1200, 240° Azimuth) Intermediate to Lower Shell Position 1.1 43.3 1.29 -30 55.9 55.9 82 Girth Weld Seam 101 - 171 Position 2.1 16.7 1.29 -30 21.5 21.5(:) 13 Lower Shell Long. Weld Seams Position 1.1 43.3 1.29 -10 55.9 55.9 102 101-142B, C (210°, 330° Position 2.1 16.7 1.29 -10 21.5 21.5(') 33 Azimuth) Lower Shell Long. Weld Seam Position 1.1 43.3 1.15 -10 49.8 49.8 90 101-142A (90° Azimuth) Position 2.1 16.7 1.15 -10 19.2 19.2(c) 28 Notes: (a) Initial RTNDT values are measured values (b) ARTPm = CF
- FF (b) M = 2 *(aj2 + A2)11 (c) RTpTs = RTNDT(u) + ARTpTs + Margin (°F)
(d) Data deemed credible per Reference 10. (e) Neutron Fluence value used for all material is the highest value from Table 5-4 for 54 EFPY. 18
PRESSURE TEMPERATURE LINIITS REPORT 6.0 Refcrcnccs
- 1. WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al.
- 2. WCAP-16142-P, Revision 1, "Reactor Vessel Closure [-lead/Vessel Flange Requirements Evaluation for Vogtle Units I and 2", Warren Bamford, et. al., February 2004.
- 3. Code of Federal Regulations, IOCFR50, Appendix H, Reactor I'esscl MaterialSurveillance ProgramRequirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 4. WCAP- 11381, GeorgiaPower Conmpany Alvin W Ibgtle Unit No. 2 Reactor l Essel Radiation Surveillance Program,L. R. Singer, April 1986.
- 5. ASTM E23 Standard Test Method Notched BarImpact Testing of Metallic h'aterials,in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 6. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
- 7. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix Q FractureToughness Criteriafor ProtectionAgainst Failure.
- 8. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, StandardPractice for ConductingSurveillance Testsfor Light- atelr Cooled luclear Power Reactor Vessels.
- 9. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Mlaterials, U.S.
Nuclear Regulatory Commission, May 1988.
- 10. WCAP- 15159, Analysis of CapsuleXFrom the Southern Nuclear Vogtle Electric Generating Plant Unit 2 Reactor Jessel RadiationSurveillance Program,T.J. Laubham, et. al., Dated March 1999. [Note that the Testing/Analysis reportsforsurveillance capsules U and Yfromn 1ogtle Unit 2 were documented under W'CAP-13007 and WCAP-14532. respectivelu.]
I1. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by AT] Consulting, March 1999. 19}}