ML14330A250
| ML14330A250 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/30/2014 |
| From: | Noronha S AREVA |
| To: | Entergy Operations, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN111401, 77-3300-001 ANP-3300, Rev 1 | |
| Download: ML14330A250 (39) | |
Text
Attachment 4 1CAN111401 ANP-3300, Revision 1, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY" November 2014
ANP-3300, Revision 1 November 2014 Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY
ANP-3300, Revision 1 November 2014 Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY Prepared by S J Noronha Reviewed by S H Mahmoud AREVA Document No.
77-3300-001 Prepared for Entergy
so A
AREVA ANP-3300, Revision 1 Copyright © 2014 AREVA.
All Rights Reserved
A AREVA ANP-3300, Revision 1 Record of Revision Revision Pages/Sections/Paragraphs No.
Changed Brief Description / Change Authorization 000 All Original Release Throughout Title pages and headings updated to reflect the revision.
Section 3.0 Updated the highest ART values and location. Updated the values in Tables 3.1 and 3.3.
Section 4.2 Minor editorial changes to clarify the description of postulated flaws in the beltline and nozzle comer.
Section 5.0 Updated the value of criticality temperature.
001 Updated the value of criticality temperature and LTOP enable temperature. Discussions updated as well.
Updated the LTOP PORV pressure limit.
Section 7.0 Updated Tables 7-1 through 7-4.
Updated Figures 7-1 through 7-3.
Updated Tables 7-5 and 7-6.
4 I.
A AREVA ANP-3300, Revision 1 Table of Contents Page RECORD OF REVISIO N........................................................................................................................
I LIST O F TABLES.................................................................................................................................
III LIST OF FIGURES...............................................................................................................................
IV 1.0 INTRODUCTIO N........................................................................................................................
1 2.0 BACKGRO UND..........................................................................................................................
1 3.0 ADJUSTED NIL-DUCTILITY TRANSITION REFERENCE TEMPERATURES........................ 3 4.0 DESIGN BASIS FOR PRESSURE-TEMPERATURE LIMITS.................................................
6 4.1 Material Properties............................................................................................................................
6 4.2 Postulated Flaws...............................................................................................................................
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 CORRECTIONS...............................................................................................
13 7.0 SUM MARY O F RESULTS........................................................................................................
14
8.0 REFERENCES
28 9.0 CERTIFICATIO N......................................................................................................................
29 ii
A AREVA ANP-3300, Revision 1 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 P Y.........................................................................................................................
4 Table 3-2: Summary of ANO-1 RV Weld Data and Adjusted Reference Temperature Results at 54 EFPY (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......................................................................
24 Table 7-6: Operational Constraints for Plant Cooldown...................................................................
24 iii
A AR EVA ANP-3300, Revision 1 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..........................................
25 Figure 7-2: Tech. Spec. Normal Cooldown P-T Limits...................................................................
26 Figure 7-3: Tech. Spec. ISLH Composite (Heatup/Cooldown) P-T Limits........................................
27 iv
A A R EVA ANP-3300, Revision 1
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 Xl, "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 600F. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Kjc). The Kic curve appears in Appendix G of ASME Code Section XI [2]. When a given material is indexed to the Kic 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
A AREVA ANP-3300, Revision 1 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 [4] and ASME Code Section Xl, Appendix G [2].
The ANO-1 reactor vessel contains both longitudinally and circumferentially oriented welds as shown in Figure 2-1. The P-T limits for ANO-1 are based on the postulation of flaws in the most limiting material.
Axial flaws are considered for forging, plates, and axial-welds. Circumferential flaws are considered in circumferential welds.
Figure 2-1: The Location and Identification of Materials Used for ANO-1 RV Page 2
"OC A
ARE VA ANP-3300, Revision 1 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 1/4A t (t - thickness of the section) and % 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 Upper Shell Plate 1, C-5120-2, with an ART value of 180 OF at the 1At wall location and an ART value of 146 OF at the %t wall location. The limiting ART values are listed in Table 3-3. Note that the limiting material is now the upper shell plate material (C-5120-2) in comparison to the weld material (WF-18) reported in the last revision of this document. This change is due to the updated initial RTNDT values for the plate material.
Page 3
Ll A
AR EVA ANP-3300, Revision 1 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
(*F)
ART ('F)
CF RTNDT 54 EFPY Fluence (n/cm 2) at 54 EFPY Margin (°F) at 54 EFPY
(*F)
Beitline Forgings Material Material Heat Cu Ni Wetted or Plates Type i ID I No.
wt%
wt%
[Note A]
[Note B]
Surface
/T
%T
/4T i 3/4 T 1/T 3/4 T
,/T 3/4 T LNBF at start of ASTM 12" thickness ASC2 AYN 131 528360 0.03 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 12" thickness A508 Cl. 2
]
8.4LNBFthcnsat start of A58STMcl2j
.5+8 84E1 69
- 2.
21 5.
N AYN 131 528360 0.03 0.70 20.0 27.5 1.45E+18 8.48E+17 3.08E+17 7.7 45 269 26.2 62.1 58.1 8.44" thickness A508 Cl. 2 LNB at LNB to A
UppCLNBFatLNBto ASTM 1
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 Weld A508 C5. 2 E{
Upper Shell llate 11 C5120-2 C5120-2 0.17 0.55 122.75 1
1.35E+19 7.90E+18 2.87E+18 114.7 80.9 63.6 63.6 179.3 145.5 Upper Shell Pl SA 533 UC5114-2 C5114-2 0.15 0.52 105.6 10 1.35E+19 7.90E+18 2.87E+18 98.6 69.6 34.0 34.0 142.6 113.6 Gr. B CI.1 i
Lower Shell Plate 1 SA 533 at 8.44" thickness Gr. B Cl 1 C5120-1 C5120-1 0.17 0.55 122.75 1
1.33E+19 I 7.78E+18 2.83E+18 114.1 80.4 63.6 J 63.6 178.7 145.0 LowerShellPlate2 SA533 C5114-1 1 C5114-1 0.15 0.52 105.6 30 1.33E+19 7.78E+18 2.83E+18 98.2 1 69.2 34.0 34.0 162.2 at 8.44" thickness Gr. B Cl. 1 7
8 3
9 6
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.
Page 4
c,~ ~
A AREVA ANP-3300, Revision 1 Table 3-2: Summary of ANO-1 RV Weld Data and Adjusted Reference Temperature Results at 54 EFPY (BAW-2308 Inputs)
Chemistry Chem.
Initial Projected ARTNDT (*F)
Margin (*F)
ART (fF)
Weld Metal Identification
[Note C]
Factor RTial 54 EFPY Fluence (n/cm2) at 54 EFPY at 54 EFPY at 54 EFPY (CF) etneed Materil i
Cu Ni Wetted T
Acepane Wire AT Beltline Welds Acceptance I etN.
w%
w%
[Note D]
[Note E]1/urac4 T
% T 1/4/ T 3/4T 1/4/ T 343/4T 1/4/ T
% T No.
Heat No.
wt%
wt%
El Surface LNBF to US WF-182-1 821T44 0.24 0.63 177.95
-84.2 1.22E+19 7.14E+18 2.59E+18 161.1 112.7 59.2 59.2 136.1 87.7 Circ. Weld US 1 to US 2-WF-18 8T1762 0.19 0.57 167.0
-48.6 1.08E+19 6.32E+18 2.29E+18 145.5 100.7 66.6 66.6 163.5 118.6 Long. Welds (2)
I US tokLS WF-112 406L44 0.27 0.59 182.55
-98.0 1.30E+19 7.60E+18 2.76E+18 168.5 118.5 60.6 60.6 131.1 81.1 Circ. Weld LS i to LS 2 WF1 Long.WF-18 W
(8T1762 0.19 0.57 167.0
-48.6 1 1.16E+19 6.79E+18 2.46E+18 148.8 103.6 66.6 66.6 166.8 121.6 Long. Welds (2) i 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.
Page 5
A AREVA ANP-3300, Revision 1 Table 3-3: Limiting Adjusted Reference Temperatures for ANO-1 RV Vessel Component Material ID 1/4T ART 3/4T ART Upper Shell Plate 1 C-5120-2 180 'F 146 oF 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 Temp.
Elastic Thermal(2)
Thermal Specific Density Thermal Modulus Expansion Conductivity,
- Heat, Conductivity E
U.
k Cp P
for Cladding Material (OF)
(106 psi)
(10-6in/in/°F)
(Btu-in/hr-ft2-OF)
(Btu/Ib-°F)
(lb/ft3)
(Btu-in/hr-ft2-°F) 70 100 150 200 250 300 350 400 450 500 550 600 650 700 29.2 29.0 28.8 28.5 28.3 28.0 27.7 27.4 27.2 27.0 26.7 26.4 25.9 25.3 7.0 7.1 7.2 7.3 7.3 7.4 7.5 7.6 7.6 7.7 7.8 7.8 7.9 7.9 282.0 283.2 283.2 283.2 282.0 280.8 279.6 277.2 274.8 272.4 270.0 266.4 262.8 259.2 0.105 0.107 0.110 0.114 0.116 0.120 0.123 0.126 0.128 0.131 0.134 0.136 0.139 0.142 490.9 490.5 489.9 489.2 488.6 487.9 487.3 486.7 486.0 485.4 484.7 484.1 483.4 482.8 103.2 104.4 108.0 111.6 115.2 117.6 121.2 124.8 127.2 130.8 133.2 135.6 139.2 141.6 Page 6
A ANP-3300, Revision 1 AR EVA 4.2 Postulated Flaws
- a.
Postulated Reactor Vessel Beltline Flaws Semi-elliptical surface flaws that are 1/4 t deep and 11/2 t long are postulated on the inside (known as 1/4 t flaw) and outside surfaces (known as % t flaw) of the reactor vessel beltline region. A longitudinal flaw is postulated in the limiting material (C-5120-2).
- 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 outlet nozzles (bounds the inlet nozzle and the core flood nozzle).
4.3 Upper Shelf Toughness A maximum value of 200 ksi*/in 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-ft2-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 IF or less), a value of 430 BTU/hr-ft2-°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.
Page 7
A ANP-3300, Revision 1 AREVA 4.6.1 Heatup Transients The following three sets of normal heatup transients were analyzed:
60 IF - 84 IF: 15 °F/hr 84 OF - 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 270 OF.
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 OF). For each of these transient cases, the fourth RCP trip was simulated by a 25 OF temperature decrease in 20 seconds. This 25 OF 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 OF:
The step cooldown transient is defined as follows:
570 OF - 280 OF: 50 OF steps with 30 minute hold periods or equivalent 280 OF - 150 OF: 25 OF 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 OF - 60 OF: 25 OF steps with 60 minute hold periods or equivalent The ramp cooldown transient is defined as follows:
570 OF - 280 OF: 100 °F/hr ramp 280 OF - 150 OF: 50 °F/hr ramp at 270 OF: DHRS initiation as described above at 255 OF: 25 OF ramp in 20 seconds (to simulate the tripping of the fourth RCP)
Page 8
A ANP-3300, Revision 1 AREVA 150 IF - 60 IF: 25 °F/hr ramp Cooldown with last RCP trip at 200 OF:
The 200 IF value was selected as an intermediate value between 255 OF and 175 OF. 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 OF 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
- Km -Krr allow SFx 1Z, where, Pallow
=
allowable pressure KR
=
reference stress intensity factor (K,,)
K
=
thermal stress intensity factor k[P
=
unit 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
A ANP-3300, Revision 1 ARE VA the fluence adjusted RTNDT. 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 (272 OF) or the RTNDT of the closure flange material + 160 OF (110 IF). Hence, the criticality limit temperature is 272 OF.
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, Kit = 33.2 + 20.734 exp [0.02 (T -
RTNDT)]
The upper shelf fracture toughness is limited to an upper bound value of 200 ksiiin 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.
Page 10
A AREVA ANP-3300, Revision 1 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, aT
=a 2T +1aT.
at Cp r.*
k(.
2 r-7.r) subject to the following boundary conditions:
at the inside surface, where r = Ri, aT
- k
= h(Tw - Tb at the outside surface, where r = Ro, aT
-=0
- where, p=
density Cp=
specific heat k =
thermal conductivity T=
temperature r=
radial coordinate t=
time h =
convection heat transfer coefficient T=
wall temperature Tb =
bulk coolant temperature Ri=
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.
Page 11
A ANP-3300, Revision 1 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:
Eax I r 2 +R R2 Trdr+JfTrdr-Tr2
[Ref. 10, Eqn(255)]
1-v r R2 -R The thermal stress distribution is then expressed by the following polynomial:
o(x) = Co + C1 (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 Co + 0.6322 CI + 0.4753 C 2 + 0.3855 C 3 )Ia For a 1/4 t outside surface flaw during heatup, Kt=
(1.043 C0 + 0.630 C1 + 0.481 C2 + 0.401 C3)O Ia 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 Kirn=
MmxRi/t where R1 =
vessel inner radius, in.
t =
vessel wall thickness, in.
For a longitudinal 1/4 -thickness x '/2 -thickness semi-elliptical surface flaw:
at the inside surface, Mm =
1.85 for 4t < 2
=
0.926 4t for 2 < 'It < 3.464 Page 12
A ANP-3300, Revision 1 AREVA
3.21 for 4t > 3.464 at the outside surface, Mr
1.77 for 4t < 2
= 0.893 4It for 2 < 4/t < 3.464
=
3.09 for 4t > 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 = cia F(a/rn) where o'
Ri/t R=
nozzle belt shell inner radius, in.
t =
nozzle belt shell wall thickness, in.
a =
flaw depth, in.
rn=
apparent radius of nozzle, in.
=
ri + 0.29rc ri =
inner radius of nozzle, in.
rc=
nozzle corner radius, in.
and F(a/rn) = 2.5 - 6.108(a/rn) + 12(a/rm) 2 - 9.1664(a/rn) 3 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
A ANP-3300, Revision 1 ARE VA 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, OF AP, psi RCP 2 AP, psi RCP 2 AP, psi RCP 2 AP, psi RCP 2 AP, psi RCP 2 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 532 °F 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. These P-T Page 14
A ANP-3300, Revision 1 AREVA limit curves are developed based on the pressure correction factors summarized in Table 6-1. The LTOP pressure limits are derived considering the RCP operational constraints for plant heatup and cooldown as provided in Tables 7-5 and 7-6, respectively.
Tables 7-5 and 7-6 represent the most conservative RCP operational constraints by which both the P-T limits and the LTOP limits remain valid. 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 272 OF. The criticality-limit P-T limits are shown in Table 7-2(b).
In BAW-10046A Rev. 2 [4], 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 OF 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 OF 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 259 OF plus any instrument/measurement uncertainty. This is 3 OF lower than the current (32 EFPY) LTOP enable temperature of 262 OF.
The LTOP pressure limit is determined as 553.8 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
A ARE'VA ANP-3300, Revision 1 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 Ihr Temperature (OF)
(psig)
(psig)
(psig) 60 557 535 513 65 557 535 513 70 557 535 513 75 557 535 513 80 557 535 513 84 557 535 513 89 557 535 513 94 557 535 513 99 557 535 513 104 557 535 513 109 557 535 513 114 559 535 513 119 563 535 513 123 569 536 513 124 570 537 513 129 579 538 513 134 590 543 513 139 604 550 515 144 620 560 520 149 638 571 526 154 659 585 534 159 680 601 544 164 698 620 557 169 719 641 571 174 741 665 588 179 766 691 607 184 793 720 629 189 823 753 652 194 856 791 680 199 893 831 711 204 933 877 746 209 978 922 785 214 1028 965 827 219 1083 1012 875 Page 16
A AREVA ANP-3300, Revision 1 Table 7-1 Tech. Spec. P-T Limits for Normal Heatup (continued)
Governing Adjusted Pressure Fluid Tmeaure at 50°F /hr at 70°F /hr at 90°F /hr Temperature 224 1144 1065 928 229 1211 1123 987 234 1285 1187 1052 239 1367 1258 1125 244 1457 1337 1207 249 1557 1423 1296 254 1683 1534 1369 259 1819 1658 1479 264 1964 1787 1611 269 2123 1927 1759 274 2297 2082 1900 279 2489 2251 2050 284 2701 2438 2216 289 2934 2645 2398 294 3024 2872 2600 299 3024 3024 2821 304 3024 3024 3024 Page 17
A AREVA ANP-3300, Revision 1 Table 7-2: Tech. Spec. Criticality Limit P-T Limits (a) Criticality Limit Determination Criticality Limit Temp. at 2500 psig during ISLH Pressure Temp.
(psig)
(OF) 2386*
269*
2574*
274" Inter olating:
2500 272
- From Table 7-4 (b) Criticality Limit P-T Limits Fluid Governing Temp Adjusted Pressure (OF)
(psig) 272 0
272 1026 274 1052 279 1125 284 1207 289 1296 294 1369 299 1479 304 1611 309 1759 314 1900 319 2050 324 2216 329 2398 334 2600 339 2821 344 3024 Page 18
A AREVA ANP-3300, Revision 1 Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown Fluid Governing Temp.
Adjusted Pressure (OF)
(psig) 60 508 65 508 70 508 75 508 80 508 85 508 90 508 95 508 100 508 105 517 110 526 115 535 120 546 123 553 125 557 130 568 135 578 140 591 145 604 150 628 155 659 160 701 165 739 170 778 175 798 185 919 190 960 193 993 198 1045 203 1103 208 1166 213 1236 218 1313 223 1399 Page 19
A AREVA ANP-3300, Revision 1 Table 7-3: Tech. Spec. P-T Limits for Normal Cooldown (continued) 228 1493 233 1598 238 1713 243 1841 248 1982 253 2125 255 2192 260 2371 265 2538 270 2538 340 2538 345 2538 350 2538 355 2538 360 2538 365 2538 370 2538 380 2538 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
A AREVA ANP-3300, Revision 1 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
A ANP-3300, Revision 1 AREVA Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve Fluid Governing Temp.
Adjusted Pressure (OF)
(psig) 60 668 65 714 70 714 75 714 80 714 84 714 89 714 94 714 99 714 104 719 105 719 109 719 114 720 119 720 124 720 129 720 134 721 139 723 144 729 149 737 154 748 159 762 164 778 169 798 174 820 179 846 184 875 189 906 194 943 199 985 204 1031 209 1082 214 1139 219 1203 224 1274 Page 22
A AREVA ANP-3300, Revision 1 Table 7-4: Tech. Spec. P-T Limits for ISLH HU/CD - Composite Curve (continued)
Fluid Governing Temp.
Adjusted Pressure (OF)
(psig) 229 1352 234 1439 239 1536 244 1646 249 1764 254 1867 259 2013 264 2189 269 2386 274 2574 279 2774 284 2995 289 3238 294 3507 299 3507 304 3507 309 3507 Page 23
A AR EVA ANP-3300, Revision 1 Table 7-5: Operational Constraints for Plant Heatup CONSTRAINT RC TEMPERATURE HEATUP RATE RCP RESTRICTIONS RC Temperature T < 840F
< 151F in any 1 hr period NA T _> 840F
<501F, 70°F or 90°F in NA any 1 hr period RC Pumps T > 3001F NA None 100 0F < T < 300OF NA
_< 3 pumps T < 100OF NA No pumps operating Table 7-6: Operational Constraints for Plant Cooldown CONSTRAINT RC TEMPERATURE COOLDOWN RATE RCP RESTRICTIONS RC Temperature T >_ 2801F
< 50°F in any 1/2 hr period NA 280°F > T > 150OF
< 251F in any 1/2 hr period NA T<150 0F
< 250F in any 1 hr period NA T > 250°F N/A None RC Pumps 250°F > T _>
100OF N/A
- 2 pumps T < 100°F N/A No pumps operating Page 24
A AR EVA ANP-3300, Revision 1 Figure 7-1: Tech. Spec. Normal Heatup and Criticality Limit P-T Limits 2000 02 0.
CL.
1500 1000
-,-Tech Spec P-T Limits for Normal HU @50 F/hr -Tech Spec P-T Limits for Normal HU @70 °F/hr
--$-Tech Spec P-T Limits for Normal HU @90 °F/hr Critical Core, P-T Limits (90 °Flhr) 2 ~
~
~
~
~
~
~
-9 4....
500 0
0 50 100 150 200 250 300 350 Temperature, OF Page 25
A ANP-3300, Revision 1 AR EVA Figure 7-2: Tech. Spec. Normal Cooldown P-T Limits
-CD, RT at 175 OF 2000
-CD, RT at 200F 1-CD, RT at 255 F
[ --
Tech Spec Curve for Normal CD
/
1500 0.
Q 1000 0~500 1 0 2 0 0
4 1
0 50 100 150 200 250 300 Temperature, OF Page 26
A AR EVA ANP-3300, Revision 1 Figure 7-3: Tech. Spec. ISLH Composite (Heatup/Cooldown) P-T Limits 2000 +-
-HU 509F/ hr
-HU 70°FI hr
-HU 90°F/ hr
-CD, RT @255 F
-CD, RT @200 F
-CD, RT @175 TF e Tech Spec ISLH HUICD Composite HU-CD Curve q/
/ I/I IL 1500 1000
/
lx r
I 500 0
0 50 100 150 200 250 300 350 Temperature, OF Page 27
I,
'., t A
ANP-3300, Revision 1 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 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 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 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 28
MI~**
A AREVA ANP-3300, Revision 1 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 X1, Appendix G, 2001 Edition with Addenda through 2002.
1ý3ý &qAwv%ý it I1+11"4 Silvester J. Noronha, Principal Engineer Component Analysis and Fracture Mechanics Date This report has been reviewed for technical content and accuracy.
11 ILI Samer H. Mahmoud, Principal Engineer Component Analysis and Fracture Mechanics Date Verification of independent review.
Timothy M. Wiger, Engineering Manager Component Analysis and Fracture Mechanics Date This report is approved for release.
David J. Skulina, Project Manager Date Page 29 1CAN111401 ANO-1 Pressurized Thermal Shock Reference Temperatures at 54 EFPY to 1CAN 111401 Page 1 of 1 ANO-1 Pressurized Thermal Shock Reference Temperatures at 54 EFPY RV Material Identification Chem. Composition Initial 54 EFPY Screening ial.Lo.t.n" Mat r.a
.p'..
Mat ria'.I.He t.N..C.wt.N.wt
.Fac or T NT(°F P ekP n cmkFlu nceFac or F)(rFt(e)rCite Factor RTD ea lece2 Fauece (RTN)
Margi Rrrs Crteio Material Locato MaeIa Material ID Heat No.
Cu wt%
Ni wt%(F) n/m) aco (F)
()
(F (F
Type LNBF at start of ASTM L2Nathickness ASTM Cl. 2 AYN 131 528360 0.03 0.70 20.0 27.5 1.13E+18 0.442 8.8 27.3 63.6 270 LNBF at start of ASTM 8.44 thickness ASTM C
AYN 131 528360 0.03 0.70 20.0 27.5 1.42E+18 0.491 9.8 27.6 64.9 270 8.44" thickness A508 C6. 2 LNBF at LNBF to ASTM Upper Shell Weld A508 CI. 2 AYN 131 528360 0.03 0.70 20.0 27.5 1.21E+19 1.053 21.1 33.3 81.9 270 Upr e 53 Upper Shell Plate 1 SA 533 C5120-2 C5120-2 0.17 0.55 122.75 1
1.34E+19 1.081 132.7 3.6 197.4 270 Gr. B Cl. 1 C
Uppr SellPlae 2 SA 533 i
UpperBShell Plate2 C5114-2 C5114-2 0.15 0.52 105.6 10 1.34E+19 1.081 114.2 34.0 158.2 270 Gr. B CI.1 I Lower Shell Plate 1
- SA 533 at 8.44" thickness Gr. B Cl. 1 C5120-1 C5120-1 0.17 0.55 122.75 1
1.32E+19 1.077 132.2 63.6 196.9 270 LowerShell Plate2 SA533 C5114-1 C5114-1 0.15 0.52 105.6 30 1.32E+19 1.077 113.8 34.0 177.8 270 at 8.44" thickness Gr. B Cl. 1 Axia Wels (
LNBF to US Linde 80 WF-182-1 821T44 0.24 0.63 177.95
-84.2 1.21E+19 1.053 187.4 59.2 162.4 300 US 1 to US 2 6
AilWls()
Linde 80 WF-18 8T1 762 0.19 0.57 167.0
-48.6 1.06E+19 1.016 169.7 66.6 187.7 270 US to LSjWF126.
Cr Wl 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 1 to LS 2 Linde 80 WF-18 8T1762 0.19 0.57 167.0
-48.6 1.14E+19 1.037 173.1 66.6 191.1 270 Axial Welds (2)
B 1
1 1
1 1
LNBF = Lower Nozzle Belt Forging, US = Upper Shell, LS = Lower Shell, Circ. = Circumferential