ML14241A241
| ML14241A241 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/30/2014 |
| From: | Noronha S AREVA |
| To: | Entergy Operations, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14241A319 | List:
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| References | |
| 1CAN081403, 43-3300-000 ANP-3300 | |
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Attachment 4 1CAN081403 ANP-3300, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY" June 2014
ANP-3300 June 2014 Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY
Controlled[>c mn ANP-3300 June 2014 Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY Prepared by S J Noronha Reviewed by D E Killian AREVA Document No.
43-3300-000 Prepared for Entergy
Con~tiofled Docum en*
A A REVA ANP-3300 Copyright © 2014 AREVA.
All Rights Reserved
onMYIled Document Record of Revision Revision Pages/Sections/Paragraphs No. Changed Brief Description / Change Authorization 000 All Original Release i i 4 4
Cont olled Document A ANP-3300 AR EVA Table of Contents Page RECO RD O F REVISIO N ........................................................................................................................... I LIST O F TABLES ................................................................................................................................... III LIST OF FIG URES ................................................................................................................................. IV 1.0 INTRO DUCTIO N ........................................................................................................................... 1 2.0 BACKG RO UND ............................................................................................................................. 1 3.0 ADJUSTED NIL-DUCTILITY TRANSITION REFERENCE TEMPERATURES ......................... 3 4.0 DESIGN BASIS FOR PRESSURE-TEMPERATURE LIMITS .................................................. 6 4 .1 Ma te ria l P ro p e rtie s ........................................................................................................................... 6 4 .2 P o stu la te d F la w s .............................................................................................................................. 7 4.3 Upper Shelf Toughness ............................................................................................ ........ 7 4.4 Uncorrected Reactor Vessel Closure Head Limits ...................................................................... 7 4.5 Convection Film Coefficient ........................................................................................................ 7 4.6 Reactor Coolant Temperature-Time Histories ............................................................................ 7 4.6.1 Heatup Transients ........................................................................................................ 8 4.6.2 Cooldown Transients .................................................................................................... 8 5.0 TECHNICAL BASIS FOR PRESSURE-TEMPERATURE LIMITS ........................................... 9 5.1 Fracture Toughness ....................................................................................................................... 10 5.2 Thermal Analysis and Thermal Stress Intensity Factor ............................................................ 11 5.3 Unit Pressure Stress Intensity Factor for Reactor Vessel Beltline Region ............................... 12 5.4 Unit Pressure Stress Intensity Factor for Reactor Vessel Nozzles ............................................ 13 6.0 PRESSURE CO RRECTIO NS ................................................................................................ 13 7.0 SUM M ARY O F RESULTS .......................................................................................................... 14
8.0 REFERENCES
............................................................................................................................ 29 9.0 CERTIFICATION ......................................................................................................................... 30 ii
Con., oied !Eocurenei A ANP-3300 AR EVA List of Tables Page Table 3-1: Summary of ANO-1 RV Forging and Plate Data and Adjusted Reference Temperature R e su lts at 54 E F PY ........................................................................................................................... 4 Table 3-2: Summary of ANO-1 RV Weld Data and Adjusted Reference Temperature Results at 54 E FPY (BAW -2308 Inputs) .................................................................................................... ...... . 5 Table 3-3: Limiting Adjusted Reference Temperatures for ANO-1 RV ................................................ 6 Table 4-1: Reactor Vessel Steel and Cladding Material Properties ..................................................... 6 Table 6-1: Limiting Location Pressure Corrections Factors for ANO-1 ................................................ 14 Table 7-1: Tech. Spec. P-T Limits for Normal Heatup ....................................................................... 16 Table 7-2: Tech. Spec. Criticality Limit P-T Limits ........................................................................... 18 Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown .................................................................. 19 Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve ........................................... 22 Table 7-5: Operational Constraints for Plant Heatup ....................................................................... 25 Table 7-6: Operational Constraints for Plant Cooldown ................................................................... 25 iii
Controled Docurent A
AREVA ANP-3300 List of Figures Page Figure 2-1: The Location and Identification of Materials Used for ANO-1 RV ................................... 2 Figure 7-1: Tech. Spec. Normal Heatup and Criticality Limit P-T Limits ........................................... 26 Figure 7-2: Tech. Spec. Normal Cooldown P-T Limits ..................................................................... 27 Figure 7-3: Tech. Spec. ISLH Composite (Heatup/Cooldown) P-T Limits ........................................ 28 iv
Controlled Docurnnt A ANP-3300 AR EVA
1.0 INTRODUCTION
This report provides Reactor Coolant Pressure Boundary (RCPB) Technical Specification Pressure-Temperature (P-T) operating limits for Arkansas Nuclear One Unit 1 (ANO-1) at 54 effective full-power years (EFPY) of operation. The P-T limits are established in accordance with the requirements of 10 CFR Part 50, Appendix G [1]. These P-T limits are generated for normal operation heatup, normal operation cooldown, inservice leak and hydrostatic test (ISLH) conditions, and reactor core operations.
These limits are expressed in the form of curves of allowable pressure versus temperature. The uncorrected P-T limits for ANO-1 were determined for 54 effective full power years (EFPY) of operation.
Pressure correction factors were determined between pressure sensor locations in the reactor coolant system (RCS) hot leg and various regions of the reactor vessel (RV). In addition, the minimum temperature for core criticality is determined to satisfy the regulatory requirements of 10 CFR Part 50, Appendix G [1].
2.0 BACKGROUND
The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The three areas of the reactor pressure vessel addressed in the present report are the beltline shell region, the reactor coolant nozzles, and the closure head flange region.
A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" [2]. This method utilizes fracture mechanics concepts and the reference temperature for nil-ductility transition (RTNDT). The RTNDT is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E208 [3]) or the temperature at which the material exhibits 50 ft-lbs absorbed energy and 35 mils lateral expansion minus 60 0 F. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (K,,). The Kic curve appears in Appendix G of ASME Code Section XI [2]. When a given material is indexed to the Klo curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Plant operating pressure-temperature limits can then be determined using these allowable stress intensity factors.
The beltline region of the reactor vessel is the most highly exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels such as SA-533, Grade B Class 1, and SA-508, Class 2 forging material used in the fabrication of the ANO-1 reactor vessel and inlet and outlet nozzles, are well characterized and documented in the literature. The Page 1
coi-i'olld Doc\ -,ueni A ANP-3300 AREVA effects of irradiation on these steels include an increase in the yield and ultimate strengths and a decrease in ductility. The most significant effect, however, is an increase in the temperature associated with the transition from brittle to ductile fracture and a reduction in the Charpy upper-shelf energy value.
Pressure-temperature limits for the ANO-1 reactor vessel are developed in accordance with the requirements of 10 CFR Part 50, Appendix G [1], utilizing the analytical methods and flaw acceptance criteria of topical report BAW-10046A, Revision 2 [41 and ASME Code Section XI, Appendix G [2].
The ANO-1 reactor vessel contains both longitudinally and circumferentially oriented welds as shown in Figure 2-1, Therefore, the P-T limits for ANO-1 are based on the postulation of both longitudinal (axial) and circumferential flaws in the most limiting axial and circumferential welds
--- *g Upp er ,,r on t fwa AT14 12 t; 529420 WF 12 1 FLStoI LSo2 W1an j Shall iong~udinaf weid). -t C512G4 _O1rfld C51 14-1
'Wedd WF 11 Figure The 2-1Location and IdentficationofMtrilVse-o AO1R Page 2
Controlled Docunent A ANP-3300 AR EVA 3.0 ADJUSTED NIL-DUCTILITY TRANSITION REFERENCE TEMPERATURES The RTNDT of the reactor vessel materials, and in turn, the pressure-temperature limits of a reactor vessel, must be adjusted to account for the effects of irradiation. The adjusted RTNDT (ART) values are calculated by adding a radiation-induced ARTNDT to the initial RTNDT plus a margin term using Regulatory Guide 1.99 Revision 2 [5] to predict the radiation induced ARTNDT values as a function of the material's copper and nickel content and neutron fluence. The projected fluence values at 54 EFPY are based on NRC approved Topical Report BAW-2241 P-A, Revision 2 [6], which complies with Regulatory Guide 1.190 [7].
The 54 EFPY 1/4 t (t - thickness of the section) and 3/4 t ART values for the ANO-1 reactor vessel beltline base and weld materials are listed in Table 3-1 and Table 3-2 respectively. These values were calculated in accordance with Regulatory Guide 1.99, Revision 2 [5]. The calculation of the ART values for the weld metals used the following information from BAW-2308 Revision 1A and 2A [8]; the initial RTNDT, the associated standard deviation and the added chemistry factor requirement. Entergy has made an exemption request to the NRC [9] to utilize BAW-2308 Revision 1A and 2A for determining the ART values for the Linde 80 weld metals for the ANO-1 unit. Table 3-3 summarizes the limiting ART values for ANO-1 used in the calculation of P-T limits.
The highest ART values for the ANO-1 reactor vessel is at the Lower Shell 1 to Lower Shell 2 longitudinal weld, WF-18, with an ART value of 166.8°F at the 1/4t wall location and an ART value of 121.6°F at the /t wall location. The limiting ART values are listed in Table 3-3.
Note that the ART values of the longitudinal welds joining Upper Shell 1 to Upper Shell 2 (WF-18) are similar to the ART values of the longitudinal welds joining Lower Shell 1 to Lower Shell 2 (WF-18). The ART values for the Lower Shell 1 to Lower Shell 2 longitudinal welds are used as the limiting ART values because they are bounding. Also note that the limiting weld for the 54 EFPY P-T limits, WF-18, is different from the limiting weld for the 32 EFPY P-T limits, WF-112, as a result of using the inputs from BAW-2308.
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Controiiid Docuni-ient A ANP-3300 AR EVA Table 3-1: Summary of ANO-1 RV Forging and Plate Data and Adjusted Reference Temperature Results at 54 EFPY Base Metal Identification Chemistry Initial Projected ARTNDT (OF) Margin (OF) ART (°F) 2
....... CF RTNOT 54 EFPY Fluence (n/cm ) at 54 EFPY at 54 EFPY (OF)
Beltline Forgings Material Material Heat Cu Ni Wetted or Plates Type ID No. wt% wt% [Note A] [Note B] Surface LNBF at start of ASTM -51 2 12" thickness A C. 2 131 528360 003 0.70 20.0 27.5 1.13E+18 5.34E+17 1.26E+17 6.1 2.6 26.5 25.9 60.1 56.0 LNBF at start of ASTM 8.44" thickness A508 Cl. 2 AYN 131 528360 0.03 0.70 20.0 27.5 1.45E+18 8.48E+17 3.08E+17 7.7 4.5 26,9 26.2 62.1 58.1 LNBF at LNBF to A.STM !
UperShlNBWel ASTM I AYN 131 528360 0.03 0.70 20.0 27.5 1.22E+19 7.14E+18 2.59E+18 18.1 12.7 31.5 28.7 77.1 68.9 Upper Shell Wel N 58 l SA 533 Upper Shell Plate 1 C5120-2 C5120-2 0.17 0.55 122.75 -10 1.35E+19 7.90E+18 2.87E+18 114.6 80.9 34.0 34.0 138.6 104.9 Gr. B Cl. 1 I-.
SA 533 -10 Upper Shell Plate 2 C5114-2 C5114-2 0.15 0.52 105.6 1.35E+19 7.90E+18 2.87E+18 98.6 69.6 34.0 34.0 122.6 93.6 Gr. B Cl. I . . ........... ......... . .......... ........... ................
.................. . .. .t. . . .. . . . .
.... 5 ..........................
Lower Shell Plate 1 SA 533 1.33E+19 7,78E+18 C5120-1 C5120-1 0.17 122.75 -10 2.83E+18 114.1 80.4 34.0 34.0 138.1 104,4 at 8.44" thickness Gr. B CI. 1 f
¢,, ^ *,z,,.*
Lower Shell Plate 2
- 005 C5114-1 C5114-1 0.15 0.52 105.6 0 1.33E+19 7.78E+-18 2.83E+18 98.2 1 69.2 34.0 34.0 132.2 103.2 at 8.44" thickness Gr, B Cl. 1 1 LNBF = Lower Nozzle Belt Forging Notes:
A. Chemistry Factor is calculated per Regulatory Guide 1.99, Revision 2 [5], Table 2 (linear interpolation allowed).
B. Initial RTNDT for the Lower Nozzle Belt Forging is a generic mean value for pre-1971 A508 Class 2 forgings manufactured by Ladish Company; Initial RTNDT values for Upper and Lower Shell Plates are measured values.
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rcF~ed Do2u (~d~
A AR EVA ANP-3300 Table 3-2: Summary of ANO-1 RV Weld Data and Adjusted Reference Temperature Results at 54 EFPY (BAW-2308 Inputs)
Chem.
Chemistry Projected ARTNDT ('F) Margin ('F) ART (°F)
Weld Metal Identification Chem. Initial at 54 EFPY at 54 EFPY at 54 EFPY
[Note C] 54 EFPY Fluence (n/cm Factor RTNDT
('F)
Material Wire Wetted Cu Ni 14T /4T '/AT 3/4T Beltline Welds Acceptance [Note D] [Note E] Surface 14T Heat No. wt% wt%
No.
LNBF to US 1.22E+19 7.14E+18 2.59E+18 161.1 112.7 59.2 59.2 136.1 87.7 WF-182-1 821T44 0.24 0.63 177.95 -84.2 Circ. Weld US I to US 2 167.0 -48.6 1.08E+19 6.32E+18 2.29E+18 145.5 100.7 66.6 66.6 163.5 118.6 WF-18 8T1762 0.19 0.57 Long. Welds (2)
....6.0..........
. ....... .6.....
US to LS 2.76E+18 168.5 118.5 60.6 1 60.6 131.1 81.1 WF-112 406L44 0.27 0.59 182.55 -98.0 1.30E+19 7.60E+18 Circ. Weld LS 1 to LS 2 1.16E+19 6.79E+18 2.46E+18 148.8 166.8 121.6 WF-18 8T1762 0.19 0.57 167.0 -48.6 103.6 66.6 ý 66.6 Long. Welds (2)
LNBF = Lower Nozzle Belt Forging US = Upper Shell LS = Lower Shell Circ. = Circumferential Long. = Longitudinal Notes:
C. Cu wt% and Ni wt% weld wire heat best-estimates.
D. Chemistry Factor is calculated per Regulatory Guide 1.99, Revision 2 [5], Table 1 (linear interpolation allowed) with a minimum of 167°F per BAW-2308 [8].
E. Initial RTNDT is a heat-specific value calculated for Linde 80 weld metals in BAW-2308 [8]; A license exemption request per 10 CFR 50.12 has been made to the NRC [9] to use these values.
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rControflec'1 DocumenK A ANP-3300 AREVA Table 3-3: Limiting Adjusted Reference Temperatures for ANO-1 RV Vessel Component Material ID '/T ART 3/.T ART Lower Shell 1 to Lower Shell 2 WF-18 166.8 121.6 Longitudinal Weld 4.0 DESIGN BASIS FOR PRESSURE-TEMPERATURE LIMITS Essential analytical parameters used in the preparation of the ANO-1 P-T limits are described below.
4.1 Material Properties Table 4-1 describes the material properties used in the development of the P-T limits for the ANO-1 unit.
The RV material properties are obtained from Section II of ASME B&PV Code [2].
Table 4-1: Reactor Vessel Steel and Cladding Material Properties Speciic DesityThermal Temp. Elastic Thermal(2) Thermal Specific DensityT Modulus Expansion Conductivity, Heat, Conductivity E c( k Cp for Cladding Material (0F) (106 psi) (10 6in/in/OF) (Btu-in/hr-ft 2-°F) (Btu/Ib- 0 F) (lb/ft3) (Btu-in/hr-ft 2-OF) 70 29.2 7.0 282.0 0.105 490.9 103.2 100 29.0 7.1 283.2 0.107 490.5 104.4 150 28.8 7.2 283.2 0.110 489.9 108.0 200 28.5 7.3 283.2 0.114 489.2 111.6 250 28.3 7.3 282.0 0.116 488.6 115.2 300 28.0 7.4 280.8 0.120 487.9 117.6 350 27.7 7.5 279.6 0.123 487.3 121.2 400 27.4 7.6 277.2 0.126 486.7 124.8 450 27.2 7.6 274.8 0.128 486.0 127.2 500 27.0 7.7 272.4 0.131 485.4 130.8 550 26.7 7.8 270.0 0.134 484.7 133.2 600 26.4 7.8 266.4 0.136 484.1 135.6 650 25.9 7.9 262.8 0.139 483.4 139.2 700 25.3 7.9 259.2 0.142 482.8 141.6 Page 6
Conhrol'-d Docuflien A ANP-3300 AREVA 4.2 Postulated Flaws
- a. Postulated Reactor Vessel Beltline Flaws 1
Semi-elliptical surface flaws that are 1/4 t deep and 11/2 t long are postulated on the inside (known as /4 t flaw) and outside surfaces (known as % t flaw) of the reactor vessel beltline region. A longitudinal flaw is postulated in the base metal and the longitudinal seam welds and a circumferential flaw is postulated in the circumferential welds.
- b. Postulated Nozzle Corner Flaw A 1/4 tNB ( tNB - the thickness at the nozzle belt) deep corner flaw is postulated on the inside surface of the reactor vessel inlet and outlet nozzles and core flood nozzle corner.
4.3 Upper Shelf Toughness A maximum value of 200 ksi'Iin is assumed for the upper shelf fracture toughness (Kc) of the reactor vessel beltline. For the nozzle forging materials, no "cut-off' limit is assumed.
4.4 Uncorrected Reactor Vessel Closure Head Limits Pressure-temperature limits for the reactor vessel head-to-flange closure region for normal operation and In-Service Leak and Hydrotest (ISLH) operation were derived for the ANO-1 reactor vessel closure head based on the K c fracture toughness curve. The Pressure-Temperature limits derived for the reactor vessel head-to-flange satisfy the minimum temperature requirements specified in Table 1 of Appendix G to 10CFR Part 50[1].
4.5 Convection Film Coefficient A value of 1000 BTU/hr-ft 2-OF was used for an effective convection heat transfer film coefficient at the cladding to base metal interface for all the times during heatup and cooldown when any Reactor Coolant Pumps (RCP) are in use. When no reactor coolant pumps are running (i.e., when the reactor coolant temperature is 250 OF or less), a value of 430 BTU/hr-ft 2-°F was used as an effective film coefficient at the cladding-to-base metal interface. The outside surface is modeled as a perfectly insulated boundary.
4.6 Reactor Coolant Temperature-Time Histories Ramped transients are modeled for normal operation heatup. Both ramped and stepped transient definitions are used for normal cooldown. The normal heatup and cooldown transients are also used to simulate the reactor coolant transients used for inservice leak and hydrostatic (ISLH) pressure testing.
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G.ontrcc---dA FJOC~lmenL A ANP-3300 AREVA 4.6.1 Heatup Transients The following three sets of normal heatup transients were analyzed:
60 OF - 84 IF: 15 °F/hr 84 IF - 570 OF; the three different ramp rates are: 50 °F/hr, 70 °F/hr and 90 °F/hr 4.6.2 Cooldown Transients For the analysis of the normal cooldown P-T limits, the cooldown transients were analyzed for a step transient as well as a ramp transient.
Initiation of the decay heat removal system (DHRS) occurs at a reactor coolant temperature of 2700 F.
DHRS initiation was modeled as a step change from 270 OF to 249 OF, with a hold at 249 OF for one minute, followed by a step temperature increase to 263 OF.
The cooldown transients were analyzed with the last Reactor Coolant Pump (RCP) tripping at three different temperatures (at 255 OF, at 200 OF, and at 175 IF). For each of these transient cases, the fourth RCP trip was simulated by a 25 OF temperature decrease in 20 seconds. This 25 (F change in temperature, at the time of the fourth RCP trip, occurs as the reactor coolant transitions from a state of RCP forced flow to one controlled by the DHRS.
Cooldown with last RCP Trip at 255 °F:
The step cooldown transient is defined as follows:
570 IF - 280 IF: 50 IF steps with 30 minute hold periods or equivalent 280 IF - 150 OF: 25 IF steps with 30 minute hold periods or equivalent at 270 OF: DHRS initiation as described above at 255 OF: 25 OF ramp in 20 seconds 150 IF - 60 OF: 25 IF steps with 60 minute hold periods or equivalent The ramp cooldown transient is defined as follows:
570 OF - 280 OF: 100 OF/hr ramp 280 IF - 150 IF: 50 °F/hr ramp at 270 IF: DHRS initiation as described above at 255 OF: 25 OF ramp in 20 seconds (to simulate the tripping of the fourth RCP)
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Controlled Documfen-m A ANP-3300 AREVA 150 OF - 60 OF: 25 °F/hr ramp Cooldown with last RCP trip at 200 IF:
The 200 OF value was selected as an intermediate value between 255 OF and 175 IF. Similar to the above transient, the fourth RCP trip was modeled by a 25 OF ramp in 20 seconds.
Acid Reducing Phase Cooldown with last RCP trip at 175 OF:
The purpose of this low temperature pump operation is to provide circulation throughout the RCS for acid reduction and control of water chemistry prior to completion of shutdown. For this special cooldown case involving an Acid Reducing Phase (last RCP trip at 175 OF), the cooldown transients are similar to the RCP trip at 255 OF. At 175 OF the fourth RCP trip is simulated by a 25 IF ramp in 20 seconds followed by a hold at 150 OF for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
5.0 TECHNICAL BASIS FOR PRESSURE-TEMPERATURE LIMITS Pressure-temperature limits are developed using an analytical approach that is in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G [2]. Additional requirements are contained in Table 1 of Appendix G to Title 10, Code of Federal Regulations, Part 50 [1].
The analytical techniques used to calculate P-T limits are based on approved linear elastic fracture mechanics methodology described in topical report BAW-10046A, Revision 2 [4]. The fundamental equation used to calculate the allowable pressure is Pilow -~
KIR
~K-KIT 1
SFxxk,,,
where, P,,,.= allowable pressure KIR = reference stress intensity factor ( K,)
K.1 = thermal stress intensity factor K*IPunit pressure stress intensity factor (due to 1 psig)
SF = safety factor For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, inlet nozzle, outlet nozzle, core flood nozzle, and closure head. In the beltline region, flaws are postulated to be present at the 1/4t and 3/4t locations of the controlling material (shell forging or circumferential weld), as defined by Page 9
Controlled Docurrent A ANP-3300 AREVA the fluence adjusted RTNOT. The reactor vessel nozzle flaws are located at the inside juncture (corner) with the nozzle shell, and the closure head flaw is located near the outside juncture with the head flange. P-T limits for the beltline and nozzle regions are calculated using a safety factor of 2 for normal operation and 1.5 for ISLH operation. The P-T limit curves presented consist of the allowable pressures for the controlling beltline flaw, inlet and outlet nozzles, core flood nozzle, and closure head, as a function of fluid temperature. These curves have been "smoothed", as necessary, to eliminate irregularities associated with the startup of the first reactor coolant pump during heatup and the initiation of decay heat removal during cooldown. After the initial determination of the P-T limit curves, location specific curves were adjusted for sensor location. No instrument error correction has been applied. The final results include the determination of a minimum/lower bound P-T curve.
The criticality limit temperature is obtained by determining the maximum required ISLH test temperature at a pressure of 2500 psig (approximately 10% above the normal operating pressure). The ISLH analysis considers the most limiting heatup and cooldown transients. The approach satisfies the requirement of Item 2.d in Table 1 of 10 CFR 50, Appendix G [1]. It requires the minimum temperature to be the larger of minimum permissible temperature for inservice system hydrostatic pressure test (259.50 F) or the RTNDT of 0 0 the closure flange material + 160OF (1 10 F). Hence, the criticality limit temperature is 259.5 F.
Various aspects of the calculation procedures utilized in the development of P-T limits are discussed below.
5.1 Fracture Toughness The fracture toughness of reactor vessel steels is expressed as a function of crack-tip temperature, T, indexed to the adjusted reference temperature of the material, RTNDT. Pressure-Temperature limits developed in accordance with ASME Code,Section XI, Appendix G [2], which permits the use of Kic fracture toughness, K= 33.2 + 20.734 exp [0.02 (T - RTNDT)]
The upper shelf fracture toughness is limited to an upper bound value of 200 ksi'in for the reactor vessel welds and shell base metal. No such "cut-off' limit is used for the fracture toughness of the reactor vessel nozzles. The crack-tip temperature needed for these fracture toughness equations is obtained from the results of a transient thermal analysis, described below.
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Oontrofld flocuret A ANP-3300 AREVA 5.2 Thermal Analysis and Thermal Stress Intensity Factor Through-wall temperature distributions are determined by solving the one-dimensional transient axisymmetric heat conduction equation, 2
aT = a-T + IaT) pC-- k(r2 + r r subject to the following boundary conditions:
at the inside surface, where r = Ri, k-k-T= h1(T1,-Tb)
(ýr at the outside surface, where r =Ro, oT
--=0 ar where, p density Cp specific heat k= thermal conductivity T= temperature r= radial coordinate t= time 1= convection heat transfer coefficient T,= wall temperature Tb = bulk coolant temperature Rj= inside radius of vessel R, =outside radius of vessel The above equation is solved numerically using a finite difference technique to determine the temperature at 17 points through the wall as a function of time for prescribed changes in the bulk fluid temperature, such as multi-rate ramp and step changes for heatup and cooldown transients.
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Controfled Document A ANP-3300 AREVA Thermal stress intensity factors are determined for a radial thermal gradient considering the through-wall temperature distribution at each solution time point. Through-wall thermal stress distributions are determined by trapezoidal integration of the following expression:
Thermal hoop stresses:
La I r-.2+R 2 i f0.
V= -I JTR' ir+fj T r -- Tr [Ref. 10. Eqn (255.)]
The thermal stress distribution is. then expressed by the following polynomial:
aT(x) = CQ + C, (x/a) + C2 (x/a) 2 + C3 (x/a) 3 ,
where, x = is a dummy variable that represents the radial distance from the appropriate (i.e., inside or outside) surface, in.
a = the flaw depth, in.
The thermal stress intensity factors are defined by the following relationships:
For a 1/4 t inside surface flaw during cooldown, K1, =(1.0359 C, + 0.6322 C 1 + 0.4753 C2 + 0.3855 C 3)Or For a 1/4 t outside surface flaw during heatup, K1, = (.1.043 C1 + 0.630 C, + 0.481 C2 + 0.401 C,3) Ja 5.3 Unit Pressure Stress Intensity Factor for Reactor Vessel Beltline Region The membrane stress intensity factor in the reactor vessel shell due to a unit pressure load is Ki"= MrXRi/t where R= vessel inner radius, in.
t= vessel wall thickness, in.
For a longitudinal %/ -thickness x 3/2 -thickness semi-elliptical surface flaw:
at the inside surface, Mrm = 1.85 for lt<2
= 0.926 ýit for 2 < */t < 3.464 Page 12
Controlle-d Documef A
AR EVA ANP-3300 ARE
= 3.21 for /t > 3.464 at the outside surface, Mm = 1.77 fornt < 2
= 0.893 4t for 2 <'It < 3.464
= 3.09 for qt > 3.464 5.4 Unit Pressure Stress Intensity Factor for Reactor Vessel Nozzles Considering a nozzle as a hole in a shell, WRC Bulletin 175 [11] presents the following method for estimating stress intensity factors for a nozzle corner flaw:
Kim = *I-K F(a/rn) where a= Ri/t R= nozzle belt shell inner radius, in.
t= nozzle belt shell wall thickness, in.
a = flaw depth, in.
r,= apparent radius of nozzle, in.
ri + 0.29r, r= inner radius of nozzle, in.
rc= nozzle corner radius, in.
and 2 3 F(a/r,) = 2.5 - 6.108(a/rn) + 12(a/r,) - 9.1664(a/r,)
6.0 PRESSURE CORRECTIONS The uncorrected P-T limits are calculated at the required locations or components in the RCS. Although both wide and low range pressure taps are located in the hot legs, they are both modeled at the same node in the thermal hydraulics model, and, therefore, only one set of location corrections is used. The uncorrected P-T limits are corrected to this single location. Location correction factors were determined for various temperatures and pump combinations. The limiting correction factors at various temperature Page 13
CronhYolled D~ocument A ANP-3300 AREVA ranges were then determined for beltline, nozzle, and closure head locations, as tabulated in Table 6-1 for ANO-1.
Table 6-1: Limiting Location Pressure Corrections Factors for ANO-1 Temperature 50-99 100-249 250-349 350-449 450-5321 Range, 'F RCP 2 AP, psi RCP 2 AP, psi RCP 2 AP, psi RCP 2 AP, psi RCP 2 AP, psi Beltline 22 0/0 109 2/1 122 2/2 116 2/2 108 2/2 Outlet Nozzle 17 0/0 71 2/0 69 2/0 47 2/2 44 2/2 RVCH 14 0/0 67 2/0 66 2/0 N/A - N/A -
Core Flood 17 0/0 106 2/1 122 2/2 116 2/2 107 2/2 Nozzle
- 1) The correction factor is used for temperatures above 5321F since the values are bounding for higher temperatures
- 2) The definition of RCP combinations used here are as follows: 0/0 - no pumps operating; 2/2 - all pumps operating; 2/0 - both pumps of loop A operating, both pumps of loop B are turned off; 2/1 - two pumps of loop A and one pump of loop B operating, one pump of loop B turned off.
7.0
SUMMARY
OF RESULTS The following is a summary of results for the ANO-1 P-T limits at 54 EFPY. The allowable pressures are corrected for location only. Correction due to instrument uncertainty is not included.
Maintaining the reactor coolant system pressure below the upper limit of the pressure-temperature limit curves ensures protection against non-ductile failure. Acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the applicable P-T limit curves. These P-T limit curves have been adjusted based on the pressure differential between point of system pressure measurement and the point in the reactor vessel that establishes the controlling unadjusted pressure limit.
The P-T limit curves provided in Figure 7-1 through Figure 7-3 have not been corrected for instrument error. The reactor is not permitted to be critical until the pressure-temperature combinations are, as a minimum, to the right of the criticality curve. The numerical values for the Technical Specification P-T curves provided in Figure 7-1 through Figure 7-3 are shown in Table 7-1 through Table 7-4. The Page 14
Controlied 1"ocu`,luent A ANP-3300 AREVA operational constraints for these curves are tabulated in Table 7-5 and Table 7-6. These Technical Specification P-T curves meet all the pressure and temperature requirements for the reactor pressure vessel listed in Table 1 of 10CFR Part 50, Appendix G[1].
The Tech. Spec. P-T limits for normal heatup for ANO Unit 1 are shown in Table 7-1. The Tech. Spec. P-T limits for normal cooldown for ANO-1 are determined by the limiting allowable pressure at every calculated temperature, as shown in Table 7-3. The Tech. Spec. P-T limits for ISLH heatup are shown in Table 7-4.
The criticality limit temperature corresponding to a pressure of 2500 psig is determined through interpolation of the ISLH heatup data in Table 7-4. As shown in Table 7-2(a), the criticality limit temperatures for ANO-1, is 259.5°F. The criticality-limit P-T limits are shown in Table 7-2(b).
In BAW-10046A Rev. 2 [41, the RCS piping and control rod drive motor tube (both parts of RCS pressure boundary) are qualified by establishing Lowest Service Temperature (LST) requirements in lieu of Appendix G analysis. The maximum allowable pressure for RCS piping during normal operation for temperatures up to 150°F is 20% of pre-service hydro-test minus the pressure correction factor [4]. It has been demonstrated that the limiting component at low temperature is the RVCH which removes the requirement to include the LST of RCS piping in the P-T limits [4]. It has also been demonstrated that a LST of 40'F for the control rod drive mechanism motor tube satisfies the ASME Code and 10 CFR Appendix G requirements.
The Low Temperature Overpressure Protection (LTOP) enable temperature for 54 EFPY is determined as 248 0 F plus any instrument/measurement uncertainty. This is 140 F lower than the current (32 EFPY) LTOP enable temperature of 262 0 F.
The LTOP pressure limit is determined as 563.8 psig. This value, after adjustment for measurement and opening uncertainty, is to be used for the ERV (Electronic Relief Valve) setpoint whenever the RCS temperature is below the LTOP enable temperature.
Page 15
C-01tro~lld Oocun-,int A ANP-3300 AREVA Table 7-1: Tech. Spec. P-T Limits for Normal Heatup Governing Adjusted Pressure Fluid at 50°F /hr at 70°F /hr at 90°F /hr Temperature (OF) (psig) (psig) (psig) 60 586 583 581 65 586 583 581 70 586 583 581 75 586 583 581 80 586 583 581 84 586 583 581 89 586 583 581 94 586 583 581 99 586 583 581 104 586 583 581 109 593 589 586 114 601 596 593 119 610 604 593 123 619 611 593 124 621 613 593 129 632 623 596 134 645 634 601 139 659 647 609 144 675 660 621 149 693 676 634 154 712 692 651 159 734 711 671 164 757 732 694 169 783 754 720 174 812 780 749 179 844 807 779 184 880 838 806 189 919 872 836 194 962 910 869 199 1009 951 905 204 1062 997 946 209 1120 1047 990 214 1185 1103 1039 219 1256 1164 1092 Page 16
Controlled DOC'Uhment A ANP-3300 AREVA Figure 7-1 Tech. Spec. P-T Limits for Normal Heatup (continued)
Governing Adjusted Pressure Fluid Temperature at 50OF /hr at 70°F /hr at 90°F /hr 224 1335 1233 1152 229 1422 1308 1218 234 1518 1391 1290 239 1624 1483 1370 244 1742 1585 1459 249 1871 1697 1557 254 2038 1845 1683 259 2215 2005 1830 264 2403 2173 1980 269 2608 2355 2142 274 2835 2555 2320 279 3024 2775 2515 284 3024 3018 2729 289 3024 3024 2966 294 - - 3024 299 - 3024 304 - 3024 Page 17
Con trolYHled Dcmnnr A ANP-3300 AREVA Table 7-2: Tech. Spec. Criticality Limit P-T Limits (a) Criticality Limit Determination
- From Table 7-4 (b) Criticality Limit P-T Limits Fluid Governing TeFup Adjusted Pressure (OF) (psig) 259.5 0 259.5 1098 264 1152 269 1218 274 1290 279 1370 284 1459 289 1557 294 1683 299 1830 304 1980 309 2142 314 2320 319 2515 324 2729 329 2966 334 3024 339 3024 344 3024 Page 18
Controlled Doculmemt A ANP-3300 AREVA Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown Fluid Fluid Governing Adjusted Temp. Pressure (TF) (psig) 60 535 65 535 70 535 75 535 80 535 85 535 90 535 95 535 100 535 105 547 110 559 115 573 120 587 123 597 125 603 130 619 135 637 140 657 145 680 150 717 155 752 160 806 165 854 170 904 175 936 180 993 185 1045 190 1103 193 1140 198 1207 203 1281 208 1364 213 1454 218 1555 223 1666 Page 19
Con rrolked Do<vurnent A ANP-3300 AREVA Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown (continued) 228 1788 233 1924 238 2074 243 2239 248 2422 253 2514 255 2532 260 2532 265 2532 270 2532 340 2532 345 2532 350 2532 355 2532 360 2532 365 2532 370 2532 380 2532 385 2538 390 2544 395 2551 400 2558 405 2565 410 2573 415 2581 420 2589 425 2598 430 2602 435 2612 440 2622 445 2633 450 2647 455 2659 460 2671 465 2684 470 2697 475 2711 480 2725 485 2739 Page 20
Uontrolve~d Docume nt A ANP-3300 AREVA Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown (continued) 490 2755 495 2770 500 2787 505 2803 510 2822 515 2840 520 2858 525 2878 530 2888 535 2918 540 2939 545 2960 550 2981 555 3002 560 3022 565 3040 570 3049 Page 21
C]ontoliled LlOCU rnnt A
AR EVA ANP-3300 AREVA Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve Fluid Governing Adjusted Temp.
Pressure (OF) (psig) 60 668 65 750 70 750 75 750 80 750 85 750 90 750 95 750 100 750 105 765 110 782 115 800 120 819 124 827 129 831 134 838 139 849 144 864 149 882 154 904 159 931 164 961 169 996 174 1035 179 1076 184 1112 189 1151 194 1195 199 1244 204 1297 209 1356 214 1421 219 1493 224 1572 229 1660 Page 22
Contfdle~d DOCuRlm~.t A
ARE VA ANP-3300 Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve (continued)
Fluid Governing Adjusted Temp. Pressure (OF) (psig) 234 1757 239 1864 244 1982 249 2112 254 2284 259 2480 264 2681 269 2897 274 3133 279 3393 284 3512 289 3512 294 3512 299 3512 304 3512 309 3512 440 3512 445 3526 450 3544 455 3560 460 3576 465 3593 470 3611 475 3629 480 3648 485 3667 490 3687 495 3708 500 3730 505 3753 510 3778 515 3801 520 3826 525 3852 530 3866 Page 23
C o troIled DrŽ..um.e.nt A ANP-3300 AREVA Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve (continued)
Fluid Governing Fluid Adjusted Temp. Pressure (OF) (psig) 535 3905 540 3933 545 3961 550 3990 555 4018 560 4044 565 4067 570 4079 Page 24
~orvroW~. iDc:L~;~n~
A ARE VA ANP-3300 AREVA Table 7-5: Operational Constraints for Plant Heatup CONSTRAINT RC TEMPERATURE HEATUP RATE RCP RESTRICTIONS RC Temperature T < 84 0F < 15°F in any 1 hr period NA T _ 841F < 501F, 701F or 90'1F in NA any 1 hr period RC Pumps T _ 250°F NA None 100°F _ T < 250°F NA <3 pumps T < 1001F NA No pumps operating Table 7-6: Operational Constraints for Plant Cooldown CONSTRAINT RC TEMPERATURE COOLDOWN RATE RCP RESTRICTIONS RC Temperature T Ž 280°F < 50°F in any 1/2 hr period NA 280"F > T > 1501F < 25 0 F in any 1/2 hr period NA T<150°F _ 25 0F in any 1 hr period NA T > 250'F N/A None RC Pumps 250°F > T ->100OF N/A *3 pumps T < 100OF N/A No pumps operating Page 25
A ANP-3300 AREVA Figure 7-1: Tech. Spec. Normal Heatup and Criticality Limit P-T Limits 2000 4-
,Tech P-T Limits for Normal HU @ 50 *Fhr Spec
-E3Tech Spec P-T Limits for Normal HU @ 70 °Flhr lil /
-i-Tech Spec P-T Limits for Normal HU @ 90 °F/hr
{
0)
U) 0.
1500 Critical Core, P-T Limits (90 °/hr)
/Z /
500 0 50 100 150 200 250 300 350 Temperature, OF Page 26
A ANP-3300 AREVA Figure 7-2: Tech. Spec. Normal Cooldown P-T Limits
- CD RT at 255 F
-CD RT at 200 TF 2000
-- CD RTat 175°F
-- "-Tech Spec Curve for Normal CD 1500 C-U) 4)
1000 L_
/1 500 0
0 50 100 150 200 250 300 Temperature, *F Page 27
Lj -
A ANP-3300 AREVA Figure 7-3: Tech. Spec. ISLH Composite (Heatup/Cooldown) P-T Limits 2500 I I
-HU, 50°F/ hr
-HU, 70°FI hr 200
-HU, 90RF2 hr 2000 -- CD, RT at 255 TF 1500 0.
U-CL (L 1000 500 0
0 50 100 150 200 250 300 350 Temperature, OF Page 28
L onf Wo ed Document A ANP-3300 AREVA
8.0 REFERENCES
- 1. Code of Federal Regulations, Title 10, Part 50, 'Fracture Toughness Requirements for Light Water Reactor Pressure Vessels, Appendix G to Part 50 - Fracture Toughness Requirements, Federal Register Vol. 78, no. 34248, June 7, 2013
- 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 2001 Edition with Addenda through 2002
- 3. ASTM Standard E 208-81, "Standard Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials, Philadelphia, PA
- 4. AREVA NP Document BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Appendix G," by H. W. Behnke et al., June 1986
- 5. US Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials,"
Regulatory Guide 1.99, Revision 2, May 1988
- 6. AREVA NP Document BAW-2241 P-A, Rev. 2, "Fluence and Uncertainty Methodologies," 2006
- 7. US Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001
- 8. AREVA NP Document BAW-2308, Rev. 1A and Rev. 2A, "Initial RTNDT of LINDE 80 Weld Materials," by K. K. Yoon, August 2005 and March 2008
- 9. ANO Exemption Letter 1CAN031403.
- 10. Timoshenko, S.P., and Goodier, J.N., "Theory of Elasticity," Third Edition, McGraw-Hill Book Company, 1970
- 11. PVRC Ad Hoc Group on Toughness Requirements, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," Bulletin No. 175, Welding Research Council, August 1972 Page 29
A ANP-3300 AREVA 9.0 CERTIFICATION Pressure-Temperature limits for the ANO-1 reactor vessel have been calculated to satisfy the requirements of 10 CFR Part 50, Appendix G using analytical methods and acceptance criteria of the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, 2001 Edition with Addenda through 2002.
lwvp'-Wý &(6 14 Silvester J. Noronha, Principal Engineer Date Component Analysis and Fracture Mechanics This report has been reviewed for technical content and accuracy.
- 000, Douglas E. Killian, Technical Consultant Date Component Analysis and Fracture Mechanics Verification of independent review.
Timothy M. Wiger, Engineering Manager Date Component Analysis and Fracture Mechanics This report is approved for release e.ýIý_\ to koc David Skulina, Project Manager Date Page 30
Attachment 5 ICAN081403 Pressurized Thermal Shock Assessment to 1CAN081403 Page 1 of I Pressurized Thermal Shock Assessment RV Material Identification] Chem. Composition Screening Chem.PeakFluence ARTNr Margin RTpCs Material Faco C. RTNDT 4 EPea Factor Criterion Material Location te SType Material ID Heat No. Cu wt% Ni wt% (°F) (n/cm 2) (°F)
LNBF at start of ASTM 27.5 1.13E+18 0.442 8.8 27.3 63.6 270 12" thickness A50820.0 LNBF at start of ASTM 8.44" thickness A508 CI.2 AYN 131 528360 0.03 0.70 20.0 27.5 1.42E+18 0.491 9.8 i 27.6 64.9 270 LNBF at LNBF to ASTM 27.5 1.21E+19 1.053 21.1 33.3 81.9 270 Upper Shell Weld A508 Cl.2 AYN 131 528360 0.03 0.70 20.0 SA 533i Upper Shell Plate 1 SAB C5120-2 C5120-2 0.17 0.55 122.75 -10 1.34E+19 1.081 132.7 34.0 156.7 270 Gr. B Cl.1 Upper Shell Plate 2 1 C5114-2 C5114-2 0.15 0.52 105.6 -10 1.34E+19 1.081 114.2 1 34.0 138.2 270 Lower Shell Plate 1I SA 533 at 8.44S thickness 1 Gr. B Cl. 1 C5120-1 C5120-1 0.17 0.55 122.75 -10 1.32E+19 1.077 132.2 34.0 156.2 270 Lower Shell Plate 2 SA 533 at 8.44" thickness Gr. B Cl. 1 C5114-1 C5114-1 0.15 0.52 1056 0 1.32E+19 1.077 1138 340 147.8 270 LNBF to US Linde 80 WF-182-1 I 821T44 0.24 0.63 177.95 -84.2 1.21E+19 1.053 187.4 59.2 162.4 300 Circ. Weld US i Welds Axial to US 2 Ai WeldS (2) 2 I Linde 80 WF-18 i 8T1762 0.19 0.57 167.0 -48.6 1.06E+19 1.016 169.7 1 66.6 187.7 270 US to LS ii' US to Linde 80 WF-112 406L44 0.27 0.59 182.55 -98.0 1.28E+19 1.069 195.1 60.6 157.7 300 Circ. Weld, LS ito LS2 Axial Welds (2)
] Linde 80 0F2 WF-18 8T1762 0.19 0.57 167.0
-48.6 1.14E+19 1.037 173.1 66.6 191.1 270 LNBF = Lower Nozzle Belt Forging, US = Upper Shell, LS = Lower Shell, Circ. = Circumferential