ML20196F607
| ML20196F607 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/22/1999 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20196F584 | List: |
| References | |
| NUDOCS 9906290257 | |
| Download: ML20196F607 (28) | |
Text
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, Attachment til to JPN-99-021 MARKED-UP TECHNICAL SPECIFICATION PAGES PRESSURE-TEMPERATURE LIMITS (JPTS-99-003)
NOTE: Deletions are shown in stdkccut, and additions are in bold.
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I l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l
DPR-59 9906290257 990622 PDR ADOCK 05000333 P PDR
JAFNPP LIST OF FIGURES Fiaures Title Pace 1
4.1-1 (Deleted) 4.2-1 (Deleted) 1 3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volume-Concentration Requirements
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3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 (Deleted) 3.6 1 Reactor Vessel Pressure - Temperature Limits Through 4-2 24 EFPY 163 Part 1 3.6 1 Reactor Vessel Pressure - Temperature Limits Through 44 32 EFPY 163a Part 2 3.6-1 Rccc::: Veccc! Precourc Temperatura Limit Thrcugh 16 EFPY 163b Part 3 DELETED 4.6-1 (Deleted) l 6.1-1 (Deleted) ;
6.2-1 (Deleted) i Amendment No.1 ', 22, ' 3, S', 72, 71, SS, OS,109,113,11 S,117,13 ^ ,137,158,1 S2, 227, 236, G47, vil
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The RT or values of the reactor vessel materials are listed on Table 3-2 of General Electric Report GE-NE-B1100732-01, " Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Capsule at 13.4 EFPY," Revision 1 (February 1998), including Errata and Addenda dated June 1999.
Insert B The second surveillance capsule containing test specimens was withdrawn in November,1996 after 13.4 EFPY. The test specimens removed were tested according to ASTM E 185-82 and the results are in General Electric Report GE-NE-B1100732-01, Revision 1 (February 1998), including Errata and Addenda dated June 1999.
Insert C Figure 3.6-1 is comprised of two parts: Part 1 end Part 2. Part 1 establishes the pressure-temperature limits for the bottom head, flange, upper vessel and beltline regions for plant operations through 24 j Effective Full Power Years (EFPY). Part 2 establishes the pressure-temperaturo limits for plant operations ;
through 32 EFPY. The curves contained in Figure 3.6-1 are developed from the General Electric Report GE-NE-B1100732-01, Revision 1 (February 1998), including Errata and Addenda dated June 1999.
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VALID TO 24 EFPY A - System Hydrotest Limit with Fuelin Vessel
- - - - - - ~ ' - - - -- --
B - Non-Nuclear Heating Limit C - Nuclear (Core Cntical) Limit Ass Ans B8H A B C 1400 '-Aan - System Hydrotest Limit -
with Fuel in Vessel- Bottom Head Aus - System Hydrotest Limit - - -
with Fuel in Vessel- Non.
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(90.312 5) (195.312 5) j (120.312.5) i 200 - --
. _ _ _ _ . . _ _ _. . ._. _ t SATURATION 0 -- --- E- l- - - - - - -- - - - - - -- --
0 50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature ( F)
Figure 3.6-1 Part 1 Reactor Vessel Pressure-Temperature Limits Through 24 EFPY Amendment No. 413,159, 163
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~
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Figure 3.6-1 Part 2 Reactor Vessel Pressure-Temperature Limits Through 32 EFPY Amendment No. 468, 163a
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Amendment No. %8,
[ .
. Attachment IV to JPN-99-021 ERRATA AND ADDENDA TO GE REPORT GE-NE-B1100732-01, REVISION 1 J
(JPTS-99-OO3) 1 l'
l l
l l
l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
. . l e
- BJB-9907 6/17/99 ATTACHMENT TO BJB-9907 ERRATA AND ADDENDA SHEET TO GE-NE-B1100732-01, REVISION 1 I
NEW YORK POWER AUTHOR DOCUMENT REVIEW STATUS STATUS NO:
1
% ACCEPTED 2 O ACCEPTED AS NOTED RESUBMITTAL NOT REQUIRED 3 O ACCEPTED AS NOTED RESUBMITTAL REQUIRED 4 O NOT ACCEPTED Permission to proceed does not constitute acceptance or appro details, ceiculacons, analysis, test methods or rnalertals developed or sek.cte ey ine meer and does not renve suosier from u comswee wit N "'8/8F 5-O / $ #1 A\/ l
'N REVIEWEDdrM u.,
k9[ h. .. f Prepared by: O Y ~_f DATE: -- b _M M B. J. Branlund, Principal Engineer Stmetural Assessment and Mitigation Verified by: Y L. J. YMy, SeMEr.gineer Structural Assessment and Mitigation Approved by: A 4 T. A. Caine, Manager Structural Assessment and Mitigation X
BJB-9907 6/17/99 ERRATA AND ADDENDA SilEET TO GE-NE-BI100732-01, REVISION 1 Page & Before Change . After Change Paragraph j Number !
Page 73, The 90 F limit applies when the head The 90 F limit applies when the head 3rd paragraph is on and tensioned, and also, when is on and tensioned. The limiting I the head is off. (When fuel has been vessel temperature is equal to the removed . . limiting RT, of the vessel materials for two conditions: 1) When fuel is in the vessel and the head is off, or 2)
When fuel has been removed from l the vessel, the head is tensioned, and the pressure is below 20 psig.
Pages 70 No non-beltline (i.e., upper vessel - Appendix C was added to the report through 75 feedwater) curve for the pressure test to include a non-beltline (i.e., upper Section 8 (Curve A) condition was provided. vessel - feedwater) curve for the Sections 8.1 - pressure test (Curve A) condition.
8.2.2 l
l l
1 l
l l
xi L
GE Nuclear Enorgy GE-NE-B1100732-01 Revision 1 APPENDIX C FITZPATRICK P-T CURVE CALCULATION METHOD FOR TIIE NON-BELTLINE (UPPER VESSEL - FEEDWATER)
CURVE A C-1
. ~-
GE Nuclear Energy GE-NE-B1100732-01
, .' Revision 1 P-T CURVE CALCULATION METHOD C.1 BACKGROUND This appendix is an addenda to Section 8.2 of the report GE-NE-B1100732-01, Revision 1 [1].
The purpose of the addenda is to add a pressure-temperature (P-T) curve to be used specifically for the non-beltline (i.e., upper vessel - feedwater nozzle) during the hydrostatic pressure test and leak test operating conditions. Note that this curve is not to be used for the flange region.
There are four vessel regions defined in the thermal cycle diagram [2] that should be monitored l l
against the P-T curve operating limits:
e Closure flange region (Region A) e Core beltline region (Region B) e Upper vessel (Regions A & B) e Lower vessel (Regions B & C) l The closure flange region hicludes the bolts, top head flange, vessel flange, and adjacent plates and welds. The P T curve methodology for the closure flange region is described in Section 8.3 of GE-NE-B1100732-01, Revision 1 [1]. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTa. The P-T curve methodology for the core beltline region is described in Sections 8.2.5 through 8.2.9 of GE-NE-B1100732-01, Revision 1 [1]. The remaining portion of the vessel (i.e., upper vessel ,
lower vessel) includes shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will be called the non-beltline region. The P-T curve methodology for the lower vessel region is described in Section 8.2.2 of GE-NE-B1100732-01, Revision 1 [1]. The P-T curve methodology for Core Not Critical Heat-up/ Cool-down curve for the upper vessel region is described in Section 8.2.3 through 8.2.4 of GE-NE-B1100732-01, Revision 1 [1].
C-2
l
- r. GE Nuclear Energy GE-NE-B1100732-01
. .' Revision 1 i Under certain conditions, the minimum non-beltline (i.e., upper vessel - feedwater nozzle) temperature can be significantly cooler than the beltline or closure flange region. These conditions can occur when Reactor Water Clean-Up is used to make up to the vessel through the
' feedwater (FW) nozzle. To account for these circumstances, individual temperature limits for the non-beltline (i.e., upper vessel - feedwater nozzle) were established. The P-T curve methodology j for pressure test curve for the non beltline (i.e., upper vessel - feedwater nozzle) region is described in the following sections of this appendix.
The P-T curves for the heat-up and cool-down operating condition apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress ofinterest is in the inner wall during ,
i cool-down and is in the outer wall during heat-up. Ilowever, as a conservative simplification, the thermal gradient stress at the 1/4T is assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because the maximum stress is used regardless of flaw location.
C.2 NON-BELTLINE REGIONS i
A discussion regarding the general methodology for the non-beltline regions is described in Section 8.2.1 of GE-NE-B1100732-01, Revision 1 [1].
As desenbed in Section 8.2.1 plots were developed for the limiting BWR/4 components; the feedwater nozzle (FW) and the control rod drive (CRD) penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables C-1 and C-2 below:
C-3
j GE Nuclear Energy GE-NE-B1100732-01
. . Revision 1 TABLE C-1 APPLICABLE BWlV4 DISCONTINUITY COMPONENTS FOR USE WITIl UPPER VESSEL - FEEDWATER CURVES A&B l
l Discontinuity Identification FW Nozzle l CRD llYD System Return Core Spray Nozzle !
Recirculation Inlet Nozzle i l
Steam Outlet Nozzle Water Level Instrumentation Nozzle '
Main Closure Flange Support Skirt
- Stabilizer Brackets Vibration Instrumentation Nozzle l Core AP and Liquid Control Nozzle Steam Water Interface l Jet Pump Instrumentation Nozzle Shell CRD and Bottom Ilead (B only)
Top IIead Nozzles (B only)
Recirculation Outlet Nozzle (B only)
TABLE C-2 APPLICABLE BWIU4 DISCONTINUITY COMPONENTS FOR USE WITII BOTTOM IIEAD/CRD CURVES A & B Discontinuity Identification CRD and Bottom IIcad Top Ilead Nozzle Recirculation Outlet Nozzle The P-T curves for the non-beltline region were conservatively developed for a large BWR/6
' (nominal inside diameter of 251 inches). The analysis is considered appropriate for FitzPatrick as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined from equations C-1 and C-2. The generic value was adapted to the conditions at FitzPatrick by using plant specific RTsm values for the reactor pressure vessel (RPV). The presence of nozzles and control rod (CRD) penetration holes of the non-beltline (i.e., upper vessel and bottom head, respectively), has made the analysis different from a shell analysis such as the beltline. This was the result of the stress concentrations and higher thermal stresses for C-4
.. .a .
, 'e
' ' GQ Ntl clear
- Energy GE-NE-B1100732-01 l Revision 1 l certain transient conditions experienced by the non-beltline region (i.e., upper vessel and the l
bottom head).
1 l The non-beltline curves are shifted based on the most limiting initial RTuor values for the i
appropriate non beltline components; the initial RTuor values are listed in Table 3-2 of i GE-NE-B1100732-01, Revision 1 [1]. The individual non-beltline (i.e., upper vessel - feedwater nozzle) curve is based on the non-beltline feedwater nozzle curve described in the next section.
l C.2.1 Pressure Test - Non-Beltline Curve A (Using Upper Vessel- Feedwater Nozzle Region) i CBI Nuclear (CBIN) modeled the 251 inch BWR/6 feedwater nozzles [3] to compute local stresses for determination of the stress intensity factor, K,. The result of that computation was K i= 143.1 ksi-in'* for an applied pressure of 1563 psig preservice hydrotest pressure. The computed value of(T-RTyor) was 154 F. The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the corner j thickness.
To evaluate the CBIN result, K iis calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code. Appendix G (Section III). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles.
A calculation of K[is shown below using the BWR/6,251 inch dimensions:
Vessel Radius, R 126.7 inches Vessel Thickness, t 6.5 inches Vessel Pressure 1563 psig Pressure stress: o = PR/t = 1563 psig
- 126.7 inches /(6.5 inches)
= 30466 psi l The factor F (a/Rn) from Figure A5-1 of WRC-175 is 1.6 where :
C-5 l
I
GE*
Nuclear Energy GE-NE-B1100732-01
,* Revision 1 a = lesser of 1/4 Tn or 1/4 Tv Tn = 71/8 inches Tv = 61/2 inches Rn = apparent radius of nozzle = Ri + 0.29 Rc Ri = actual inner radius of nozzle = 6 inches Rc = nozzle radius (nozzle corner radius) = 4.0 inches l Thus, a/Rn = 1.63/6.94 = 0.23 and the ratio of K, around the feedwater nozzle to the membrane stress * (na)'8 at places with no geometric discontinuity is 1.6.
l Including the safety factor of 1.3, the stress intensity factor, Ki , is 1.3 o (na)
- F(a/Rn):
1 Nominal Ki = 1.3
- 30.466 * (n* 1.63)'** 1.6 = 143 ksi-in
The method to solve for (T-RTsor) for a specific K iis based on the curve in Figure G-2210-1 in ASME Appendix G:
(T-RTuor) = 1n [(Ki - 26.78)! 1.223] / 0.0145 - 160 (T-RTsor) = 1n [(l43 - 26.78) /1.223] / 0.0145 - 160 (T-RTuor) = 154*F The generic pressure test P-T curve was generated by scaling 143 ksi-in'* by the nominal pressures and calculating the associated (T-RTsor):
Pressure Test Feedwater Nozzle Ki and (T - RTsor) as a Function of Pressure Nominal Pressure Ki (T-RTuor)
(psig) (ksi-in) ( F) 1563 143 154 l
1400 128 145 1200 110 131 1000 92 114 800 73 91 600 55 56 400 37 -16 l C-6 t
L
l GE,Nu, clear Energy GE-i.E-B1100732-01
. . Revision 1 j l
l The P-T curve is dependent on the Ki value calculated, which is proportional to the stress and the l crack depth according to the relationship:
K, oc a (na)
The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, Ki is proportional to R/(t).
The generic curve value of R/(t), based on the BWR/6,251-inch feedwater nozzle dimensions is:
Generic: R/(t) = 126.7 / (6.5)" = 49.7 inch , (C-1) where t is the nominal vessel thickness. The FitzPatrick specific vessel shell dimensions applicable to the feedwater nozzle location are R = 110 inches and t = 5.9 inches nominal.
FitzPatrick specific: R/(t) = 110 / (5.9) = 45 inch (C-2)
Since the generic value of R/(t) is greater than that for FitzPatrick, the generic P-T curve is conservative when applied to the FitzPatrick feedwater nozzle.
The highest RTwur for the feedwater region component (nozzle #N2, the Recirculation Inlet Nozzle) at FitzPatrick is 30 F as shown in Table 3-2 of GE-NE-B1100732-01, Revision 1 [1].
The generic pressure test P-T curve is applied to the FitzPatrick feedwater nozzle cun e by shifting the P vs. (T-RTypr) values above to reflect the RTsor value of 30 F. This non-beltline (i.e., upper vessel - feedwater nozzle) P-T curve is tabulated in Table C-3. The only difference between Table C-3 and Table 8-1 is the addition of the non-beltline (i.e., upper vessel - feedwater nozzle) curve tabulation. Note that this curve does not apply to the flange region. Since non-beltline curves are not influenced by irradiation, the curve is applicable for any EFPY.
C-7
. ..'.e G5 Guclear*
Energy GE-NE-B1100732-01 Revision 1 l TABLE C-3. FitzPatrick P-T Curve Values for 32 EFPY Replaces Table 8-1 of Report GE-NE-B11-00732-01, Revision 1 ;
Required Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A I
BOTTOM NON- RPV & BOTTOM RPV & RPV &
PRESSURE IIEAD BELTLINE 32 EFPY llEAD 32 EFPY 32 EFPY BELTLINE BELTLINE BELTLINE l CURVE A CURVE A CURVE A CURVE B CURVE B CURVE C l (PSIG) (F) (F) ( F) (F) ( F) (F) 0 68.0 68.0 90.0 68.0 90.0 90.0 10 68.0 68.0 90.0 68.0 90.0 90.0 l 20 68.0 68.0 90.0 68.0 90.0 90.0 30 68.0 68.0 90.0 68.0 90.0 90.0 40 68.0 68.0 90.0 68.0 90.0 94.5 50 68.0 68.0 90.0 68.0 90.0 105.2 60 68.0 68.0 90.0 68.0 90.0 113.9 70 68.0 68.0 90.0 68.0 90.0 121.1 80 68.0 68.0 90.0 68.0 90.0 127.4 90 68.0 68.0 90.0 68.0 92.7 132.7 100 68.0 68.0 90.0 68.0 97.5 137.5 110 68.0 68.0 90.0 68.0 101.9 141.9 ,
120 68.0 68.0 90.0 68.0 106.1 146.1 130 68.0 68.0 90.0 68.0 110.1 150.1 140 68.0 68.0 90.0 68.0 113.6 153.6 150 68.0 68.0 90.0 68.0 116.8 156.8 160 68.0 68.0 90.0 68.0 119.8 159.8 170 68.0 68.0 90.0 68.0 122.8 162.8 180 68.0 68.0 90.0 68.0 125.6 165.6 190 68.0 68.0 90.0 68.0 128.2 168.2 200 68.0 68.0 90.0 68.0 130.6 170.6 210 68.0 68.0 90.0 68.0 132.9 172.9 220 68.0 68.0 90.0 68.0 135.2 175.2 230 68.0 68.0 90.0 68.0 137.4 177.4 240 68.0 68.0 90.0 68.0 139.4 179.4 250 68.0 68.0 90.0 68.0 141.4 181.4 260 68.0 68.0 90.0 68.0 143.3 183.3 270 68.0 68.0 90.0 68.0 145.1 185.1 280 68.0 - 68.0 90.0 68.0 147.0 187.0 290 68.0 68.0 90.0 68.0 148.7 188.7 300 68.0 68.0 90.0 68.0 150.3 190.3 310 68.0 68.0 90.0 68.0 152.0 192.0 312.5 68.0 68.0 90.0 68.0 152.3 192.3 312.5 68.0 68.0 120.0 68.0 152.3 208.7 320 68.0 68.0 120.0 68.0 153.5 208.7 330 68.0 68.0 120.0 68.0 155.1 208.7 f
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Nuclear Energy GE-NE-B1100732-01 \
",* Revision 1 TABLE C-3. FitzPatrick P-T Curve Values for 32 EFPY Replaces Table 8-1 of Report GE-NE-B11-00732-01, Revision 1 Required Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A I
l l
BOTTOM NON- RPV & BOTTOM RPV & RPV & )
PRESSURE IIEAD BELTLINE 32 EFPY HEAD 32 EFPY 32 EFPY l BELTLINE BELTLINE BELTLINE CURVE A CURVE A CURVE A CURVE B CURVE B CURVEC (PSIG) (F) (F) (F) (F) (F) ( F) 340 68.0 68.0 120.0 68.0 156.6 208.7 350 68.0 68.0 120.0 68.0 158.0 208.7 360 68.0 68.0 120.0 68.0 159.4 208.7 370 68.0 68.0 120.0 68.0 160.8 208.7 380 68.0 68.0 120.0 68.0 162.1 208.7 390 68.0 68.0 120.0 68.0 163.4 208.7 400 68.0 68.0 120.0 68.0 164.7 208.7 i
410 68.0 68.0 120.0 68.0 166.0 208.7 420 68.0 68.0 120.0 68.0 167.2 208.7 430 68.0 68.0 120.0 70.3 168.4 208.7 440 68.0 68.0 120.0 73.2 169.6 209.6 450 68.0 68.0 120.0 76.1 170.7 210.7 460 68.0 68.0 120.0 78.8 171.8 211.8 470 68.0 68.0 120.0 81.5 172.9 212.9 480 68.0 68.0 120.0 84.0 174.0 214.0 490 68.0 68.0 120.0 86.5 175.1 215.1 500 68.0 68.0 120.0 88.8 176.1 216.1 510 68.0 68.0 120.0 91.1 177.1 217.1 520 68.0 68.0 120.0 93.3 178.1 218.1
- 530 68.0 68.5 120.0 95.5 179.1 219.1 540 68.0 71.3 120.0 97.6 180.1 220.1 i 550 68.0 74.1 120.0 99.6 181.1 221.1 560 68.0 76.7 122.0 101.5 182.0 222.0
- 570 69.5 79.2 125.1 103.5 182.9 222.9 580 71.8 81.7 128.0 105.3 184.6 224.6 590 74.0 84.0 130.8 107.1 186.3 226.3 600 76.1 86.3 133.6 108.9 187.9 227.9 610 78.2 88.5 136.2 110.6 189.5 229.5 620 80.2 90.6 138.7 112.3 191.1 231.1 630 82.1 92.7 141.1 113.9 192.6 232.6 l 640 84.0 94.7 143.5 115.5 194.1 234.1 650 85.9 96.7 145.8 117.1 195.6 235.6 l 660 87.7 98.6 148.0 118.6 197.0 237.0 670 89.4 100.4 150.1 120.1 198.4 238.4 680 91.1 102.2 152.2 121.6 199.8 239.8 690 92.8 104.0 154.2 123.0 201.1 241.1 C-9 1
I L
{ . . . . o-GE,Nyglegr Energy GE-NE-B1100732-01
. . Revision 1 TABLE C-3. FitzPatrick P-T Curve Values for 32 EFPY Replaces Table 8-1 of Report GE-NE-B11-00732-01, Revision 1 i Required Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A BOTTOM NON- RPV & BOTTOM RPV & RPV & I PRESSURE HEAD BEL.TLINE 32 EFPY HEAD 32 EFPY 32 EFPY BELTLINE BELTLINE BELTLINE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) ( F) (F) (F) (F) (F) (F) 700 94.4 105.7 156.1 124.4 202.4 242.4 710 96.0 107.4 158.0 125.8 203.7 243.7 720 97.6 109.0 159.9 127.1 205.0 245.0 730 99.1 110.6 161.7 128.4 206.3 246.3 740 100.6 112.2 163.4 129.7 207.5 247.5
-750 102.0 113.7 165.1 131.0 208.7 248.7 760 103.5 115.2 166.8 132.3 209.9 249.9 770 104.8 116.7 168.4 133.5 211.1 251.1 780 106.2 118.1 170.0 134.7 212.2 252.2 790 107.6 119.5 171.6 135.9 213.3 253.3 800 108.9 120.9 173.1 137.0 214.4 254.4 l 810 110.2 122.2 174.6 138.2 215.5 255.5 820 111.4 123.5 176.0 139.3 216.6 256.6 830 112.7 124.8 177.5 140.4 217.7 257.7 840 113.9 126.1 178.9 141.5 218.7 258.7 850 115.1 127.3 180.2 142.6 219.7 259.7 860 116.3 128.6 181.6 143.6 220.8 260.8 870 117.5 129.8 182.9 144.7 221.8 261.8 880 118.6 130.9 184.2 145.7 222.7 262.7 l 890 119.7 132.1 185.5 146.7 223.7 263.7 900 120.8 133.3 186.7 147.7 224.7 264.7 910 121.9 134.4 187.9 148.7 225.6 265.6 920 123.0 135.5 189.1 149.7 226.5 266.5 l 930 124.0 136.6 190.3 150.6 227.5 267.5 l- 940 125.1 137.7 191.5 151.6 228.4 268.4 950 126.I 138.7 192.6 152.5 229.3 269.3 l 960 127.1 139.8 193.7 153.4 230.1 270.1 970 128.! 140.8 194.9 154.3 231.0 271.0 980 129.1- 141.8 195.9 155.2 231.9 271.9 990 130.0 142.8 197.0 156.1 232.7 272.7 1000 131.0 143.8 198.1 157.0 233.6 273.6 I l 1010 131.9 144.7 199.1 157.8 234.4 274.4 1020 132.9 145.7 200.1 158.7 235.2 275.2 1030 133.8 146.6 201.1 159.5 236.0 276.0 1040 134.7 147.6 202.1 160.4 236.8 276.8 i 1050 135.6 148.5 203.1 161.2 237.6 277.6 C - 10 i I
1 L
m . ,. o GE Nuclear Energy GE-NE-B1100732-01
- ',* Revision 1 TABLE C-3. FitzPatrick P-T Curve Values for 32 EFPY Replaces Table 8-1 of Report GE-NE-Bil-00732-01, Revision 1 Required Temperatures at 100 F/hr for Curves B & C and 20 F/hr for Curve A BOTTOM NON- RPV & BOTTOM RPV & RPV &
PRESSURE liEAD BELTLINE 32 EFPY 11EAD 32 EFPY 32 EFPY BELTLINE BELTLINE BELTLINE CURVE A CURVE A CURVE A CURVE B CURVEB CURVEC (PSIG) (F) (F) ( F) (F) (F) (F) 1060 136.5 149.4 204.1 162.0 238.4 278.4
-1070 137.3 150.3 205.0 162.8 239.2 279.2 1080 138.2 151.2 206.0 163.6 239.9 279.9 1090 139.0 152.0 206.9 164.4 240.7 280.7 1100 139.9 152.9 207.8 165.1 241.4 281.4 1110 140.7 153.7 208.7 165.9 242.2 282.2 :
1120 141.5 154.6 209.6 166.7 242.9 282.9 !
1130 142.3 155.4 210.5 167.4 243.6 283.6 )
1140 143.1 156.2 211.4 168.1 244.4 284.4 1150 143.9 157.0 212.2 168.9 245.1 285.1 l 1160 144.7 157.8 213.1 169.6 245.8 285.8 i i
1170 145.5 158.6 213.9 170.3 246.5 286.5 1180 146.2 159.4 214.7 171.0 247.2 287.2 1190 147.0 160.2 215.6 171.7 247.8 287.8 1200 147.7 160.9 218.8 172.4 250.6 290.6 1210 148.5 161.7 219.6 173.1 251.2 291.2 I 1220 149.2 162.4 220.4 173.8 251.9 291.9 1230 149.9 163.2 221.1 174.5 252.5 292.5 1240 150.6 163.9 221.9 175.1 253.2 293.2 1250 151.3 164.6 222.7 175.8 253.8 293.8 1260 152.0 165.4 223.4 176.5 254.5 294.5 1270 152.7 166.1 224.2 177.1 255.1 295.1 1280 153.4 166.8 224.9 177.8 255.7 295.7 1290 154.1 167.5 225.6 178.4 256.3 296.3 1300 154.8 168.1 226.3 179.0 256.9 296.9 1310 155.4 168.8 227.0 179.6 257.5 297.5 1320 156.1 169.5 227.7 180.3 258.2 298.2 1330 156.8 170 2 228.4 180.9 258.7 298.7 1340 157.4 170.8 229.1 181.5 259.3 299.3 1350 158.1 171.5 229.8 182.1 259.9 299.9 1360 158.7 172.1 230.5 182.7 260.5 300.5 1370 159.3 172.8 231.1 183.3 261.1 301.1 1380 159.9 173.4 231.8 183.9 261.7 301.7 1390 160.6 174.0 232.5 184.5 262.2 302.2 3
1400 161.2 174.7 233.1 185.0 262.8 302.8 C - ll t
p ..,.:c-GQNyclear Energy . GE-NE-B1100732-01 Revision 1 l
C.6 REFERENCES
[1] T. J. Griesbach," Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 F Capsule at 13.4 EFPY," GE-NE, San Jose, CA, February 1998, I
(GE-NE-B1100732-01, Rev.1). l l
[2] GE Drawing #729E762, Revision 0," Reactor Thermal Cycles," GE-APED, San Jose, l
CA.
J
[3] Appendix Cl " Analysis of the Nozzle to Shell Junction Region for 10" Welded Thennal I Sleeve Nozzles,". CBI Nuclear Company, (GE VPF# 3521-415-6).
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l C - 12 l
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