ML043080360
| ML043080360 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/01/2004 |
| From: | NRC/NRR/DLPM |
| To: | |
| References | |
| TAC MC2534 | |
| Download: ML043080360 (9) | |
Text
INDEX BASES SECTION INSTRUMENTATION (Continued)
Remote Shutdown Monitoring Instrumentation and Controls...........................................
Accident Monitoring Instrumentation......................
Source Range Monitors....................................
Traversing In-Core Probe System..........................
3/4.3.8 DELETED..................................................
3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.
Figure B3/4 3-1 Reactor Vessel Water Level.............
3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION..............
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.....................................
3/4.4.2 SAFETY/RELIEF VALVES.....................................
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................
Operational Leakage......................................
3/4.4.4 CHEMISTRY................................................
3/4.4.5 SPECIFIC ACTIVITY........................................
3/4.4.6 PRESSURE/TEMPERATURE LIMITS..............................
Table B3/4.4.6-1 Reactor Vessel Toughness..............
Figure B3/4.4.6-lFast Neutron Fluence (E>lMev) at (1/4)T as a Function of Service life..............
PAGE B
B B
B B
B B
B 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3-5 3-5 3-5 3-5 3-7 3-7 3-9 3-9 B 3/4 4-1 B 3/4 4-2 B
B B
B B
B 3/4 3/4 3/4 3/4 3/4 3/4 4-3 4-3 4-3 4-4 4-5 4-7 B 3/4 4-8 Table B3/4.4.6-2 Numeric Values for Pressure/Temperature Limits....
....... B 3/4 4-9 I HOPE CREEK xviii Amendment No. 157
Figure 3.4.6.1-1 Hydrostatic Pressure and Leak Tests PressurelTemperature Limits - Curve A 1,200 I
I I
I I
I I
I I 'I I
Y J I I
-1 _ _ I I
I 1,100 1,000
--77 II
-6 a.
n 0
LU on w
0 o-0:
z
_j enCn LU W
ILt 900 800 700 600 500 400 300 200 100 -
0 I
I I
I I
I I
I I--i I
I I
I I
I I
Boltup __.
I 79OF I
0-i I 1 i
i i
i i i i I.
i I I I I I
I I I I I
Beltline
- - - Bottom Head
-Upper Vessel I
I I
I 0
50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
All system leakage and hydrostatic pressure tests performed during the service life of the pressure boundary in compliance with ASME Code Section XI.
This figure is valid for 32 EFPY of operation.
HOPE CREEK 3/4 4-23 Amendment No. 157
Figure 3.4.6.1-2 Non-Nuclear Heatup and Cooldown PressurelTemperature Limits - Curve B 1,200 1,100 1,000 0.
0 U,
U, 0I--
IL, w
z U,trw 0.
900 800 700 600 500 400 300 j 1111:501111up L Bottom Head 79"F Upper Vessel 200 100 0
0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( 0F)
All heatup and cooldowns that are performed when the reactor is not critical at the normal heatup and cooldown rate.
This figure is valid for 32 EFPY of operation.
HOPE CREEK 3/4 4-23a Amendment No. 157
Figure 3.4.6.1-3 Core Critical Heatup and Cooldown Pressure/Temperature Limits - Curve C 1,200 1,100 1,000 cm 0
(0 i-C,
'U
'U 0,
'U Ix 900 800 700 600 500 400 300
_~~~~~~~
I I lo
_____=__t____
- v r_
___Min__imu___m s__
= ~~CIt===ica
===lit===y
I_
wi__th I__
Noma
_Wate_____r
___vel-__
_===
==
_
88==
1r==F 200 100 0
0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
All heatup and cooldowns that are performed when the reactor is critical at the normal heatup and cooldown rate.
This figure is valid for 32 EFPY of operation.
HOPE CREEK 3/4 4-23b Amendment No. 157
REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section (3.9) of the UFSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
Specifically the average rate of change of reactor coolant temperature during normal heatup and cooldown shall not exceed 1000F during any 1-hour period.
The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI, Appendix G and ASME Code Cases N-588 and N-640.
The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in UFSAR Chapter 5, Paragraph 5.3.1.5, "Fracture Toughness."
Tabulated values for the P-T curves are shown in Table B 3/4.4.6-2.
The reactor vessel materials have been tested to determine their initial RTNDT.
The results of some of these tests are shown in Table B 3/4.4.6-1.
Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTUDT.
Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations-of Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Material".
The pressure/
temperature limit curves, Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3, includes an assumed shift in RTUT for the end of life fluence.
The fluence in Bases Figure B 3/4.4.6-1 was determined using methodology described in NRC-approved General Electric Nuclear Energy Licensing Topical Report NEDC-32983P-A.
This methodology is consistent with the guidance in Regulatory Guide 1.190, Rev. 0, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."
The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, irradiated flux wires installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift.
The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 shall be adjusted, as required, on the basis of the flux wire data and recommendations of Regulatory Guide 1.99, Rev. 2.
HOPE CREEK B 3/4 4-5 Amendment No. 157
BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS BELTLINE WELD SEAM I.D.
COMPONENT OR MAT'L TYPE HEAT/SLAB OR HEAT/LOT HIGHEST ARTUDT +
CU(%)
Ni(%)
RTNDT(°F)
MARGIN (OF)
PREDICTED EOL UPPER SHELF (FT-LBS)
MAX.
EOL RTDT (
F)
Plate SA-533 GR B CL.1 5K3025-1
.15 0.71
+19 56 66 75 Weld Vert. seams for shells 4&5 D53040/
1125-02205 0.081 0.611
-30 78 118 48 NOTE:
- These values are given only for the benefit of calculating the end-of-life (EOL) RTNDT.
NON-BELTLINE COMPONENT Shell Ring Connected to Vessel Flange Bottom Head Dome Bottom Head Torus LPCI Nozzles(l)
Top Head Torus Top Head Flange Vessel Flange Feedwater Nozzle Weld Metal Closure Studs MT'L TYPE OR WELD SEAM I.D.
SA 533, GR.B, C1.1 HEAT/SLAB OR HEAT/LOT All Heats SA 533, SA 533, SA 508, SA 533, SA 508, SA 508, SA 508, All RPV SA 540, GR.B, C1.1 GR.B, C1.1 C1.2, GR.B, C1.I C1.2 C1.2 C1.2 Welds GR.B, 24 All All All All All All All All All Heats Heats Heats Heats Heats Heats Heats Heats Heats HIGHEST REFERENCE TEMPERATURE RTNDT (0F)
+19
+30
+30
-20
+19
+10
+10
-20 0
Meet 45 ft-lbs & 25 mils lateral expansion at +10°F (1) The design of the Hope Creek vessel 1/4T of the vessel thickness of 3.3 RTNDT of +29 0F.
results in these nozzles experiencing a predicted EOL fluence at x 1017 n/cm2.
Therefore, these nozzles are predicted to have an EOL I
HOPE CREEK B 3/4 4-7 Amendment No.157
8 7
0 x
A E
a, 0
C) z 6
5 4
3 2
1 4 *
- I 0
5 10 15 20 25 30 35 40 Service Life, Years*
LOWER -
INTERMEDIATE SHELL FAST NEUTRON FLUENCE (E>1 MeV)
AT 1/4 T AS A FUNCTION OF SERVICE LIFE*
Bases Figure B 3/4.4.6-1
- At 80% capacity factor (40 years = 32 EFPY)
HOPE CREEK B 3/4 4-8 Amendment No.157
BASES TABLE B 3/4.4.6-2 Numeric Values for Pressure/Temperature Limits Figure 3.4.6.1-1, Curve A Bottom Head Temperature Pressure
(°F)
(psig) 79 0
79 929 88 1040 90 1068 92 1097 94 1126 96 1157 98 1190 100 1223 Bottom Head Temperature Pressure (oF)
(psig) 79 0
79 606 88 690 92 732 96 778 100 827 104 881 108 939 112 1002 116 1070 120 1144 124 1224 Upper Vessel Temperature Pressure
(°F)
(psig) 79.0 0
79.0 292 118.0 292 118.0 925 123.0 996 128.0 1074 133.0 1161 138.0 1257 Figure 3.4.6.1-2, Curve B Upper Vessel Temperature Pressure (OF)
(psig) 79.0 0
79.0 50 79.0 75 79.0 90 79.0 100 79.0 125 79.8 175 86.6 202 90.6 220 96.6 250 98.4 260 102.6 285 103.7 292 148.0 292 148.0 740 148.0 745 148.0 750 151.6 830 155.8 910 159.7 990 163.3 1070 165.5 1150 167.5 1230 Beltline Temperature Pressure (0 F)
(psig) 79.0 0
79.0 691 88.0 743 93.0 777 98.0 814 103.0 855 108.0 900 113.0 950 118.0 1,005 123.0 1,065 128.0 1,133 133.0 1,207 Beltline Temperature Pressure (0 F)
(psig) 79.0 0
79.0 416 88.0 455 93.0 480 98.0 508 103.0 538 108.0 572 113.0 610 118.0 651 123.0 697 128.0 747 133.0 803 138.0 864 143.0 932 148.0 1,008 153.0 1,091 158.0 1,183 163.0 1,284 HOPE CREEK B 3/4 4-9 Amendment No. 157
BASES TABLE B 3/4.4.6-2 (continued)
Numeric Values for Pressure/Temperature Limits Figure 3.4.6.1-3, Curve C Temperature Pressure (OF)
(psig) 88.0 0
88.0 50 88.0 75 88.0 90 92.0 100 103.4 125 119.8 175 126.6 202 130.6 220 136.6 250 138.4 260 142.6 285 143.7 292 188.0 292 188.0 740 188.0 745 188.0 750 191.6 830 195.8 910 199.7 990 203.3 1070 205.5 1150 207.5 1230 HOPE CREEK B 3/4 4-10 Amendment No. 157