ML13330A035
| ML13330A035 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/24/1980 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13311A235 | List: |
| References | |
| TAC-55027, TAC-59980, NUDOCS 8004240395 | |
| Download: ML13330A035 (8) | |
Text
ENCLOSURE 1 3.1.3 COrnBINE0 HEATUP COLOD0 N AND PRESSURE LIMIiTATIONS Applicability:
Applies to heatup and cooldown of.the reactor coolant system.
Objective:
To maintain the structural integrity of the reactor coolant.
system throughout the lifetime of the plant.
Specification:
A. Reactor pressure and heatuo and cooldovin of the reactor coolant systen during the first 16 years of equivalent full power operation shall be limited in accordance with Figures 3.1.3a and 3.1.3b. Thereafter, limits shall be based on neutron exposure equivalent to not less than 16 years of full power operation, and Figures 3.1.3a and 3.1.3b shall be updated accordingly.
B. Figures 3.1.3a and 3.1.3b shall be updated in accord ance with the following criteria and procedures:
(1) The methods of Appendix G, "Protection Against Nonductile Failure", to Section III of the ASME Boiler and Pressure Vessel Code shall be used to obtain the allowable pressure-temperature rela tionships for the reactor coolant system.
(2) The curves in Figure 3.1.3c shall be used in predicting the reference nil-ductility tempera ture increase, ARTNDT, unless measuremehts on the irradiation specimens show ART, s greater than those predicted by the curv ', in which case a new curve having the same sl'ope as the original shall be constructed.
C. The pressurizer heatup rate of 100 0F/hour and coodown rate of 195 0F/hour shall not be exceeded.
D. The reactor shall not be brought to a critical condi tion until the pressure-temperature state is to the right of the criticality limit line as shown in Figures 3.1.3a.
Basis:
The initial Reference Nil Ductility Temperature (RT,0,.) for all reactor vessel material based on Charpy V-notch" datazdrop weight tests, and conservative estimates* is 82'F or less. The RT at the 1/4 thickness location (location of Appendix G re Kence flaw tip) increases as a.function of cumulative neutron exposure up to approximately 2400F for the core region of the reactor vessel after 30 years of operation.
- NRC Standard Review Plan Branch Technical Position.MTES 5-2.
8 0 0 42t40 5
A sixteen (16) equivalent full power year service Deriod was chosen for the operational limits given in this specification because at the end of this period the limiting RT of the reator vessel at the 1/4 thickness location is a roxiiately 217 F in the core region. This RT is at least 50O F above the RTNDT of all other regions in L primary reactor coolant system..
The highest RTND of the core region material is determined by adding the radiation induced ARTNOT for the applicable timc period to the original RT shown in the Table 3.1.3.1.
The fast neutron DT (E R 1Mev) fluence at 1/4 thickness and 3/4 thickness vessel locations is given as a function of full power service life in Figure 3.1.3d. Using the applicable fluence at the end of the year period and the copper content of the material in question, the ART is obtained from Figure 3.1.3c.
NOT Values of ART, may continue to be determined in this manner unless measurVInts on the irradiation specimens show RT s
greater than those predicted by the curves for the equivai t capsule exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from non-mandatory Appendix G in Section III of the AS-E Boiler and Pressure Vessel Code, and discussed in detail in Reference 1.
The results of these calculations are provided in Reference 2.
The design heatup nd cooldown rates for the pressurizer are 1000F:/hour and 200 F/hour, respectively.
The straight line portion of the criticality limit given in Figures 3.1.3a is at the-minimum permissable temperature for the 2485 psig in-service hydrostatic test as required by Appendix G to 10CFR Part 50. The curved portion of the criticality limit is shifted 40oF to the right of the heatup curve as required by Appendix G to 10CFR Part 50.
References:
(1) "Pressure Temperature Limits" Section 5.3.2 of Standard Review Plan, NUREG-751087, 1975.
(2) S. E. Yanichko, et al, "Analysis of Capsule F from the Southern California Edison Company San Onofre Reactor Vessel Radiation Surveillance Program", WCAP 9520, May 1979..nePormW
2500 2400 2300 E-T L TEST LUIM IT 2200
.9 j
2100 MATERI AL PROPERTY
/T
, f.r To 2
CONTROLLING MA TER IAL: INTERMEDIATE SHELL 2000 COPPER CONTENT:
w0.I8 WT PHOSPHORUS CONTENT: 0.014 WT?.
1900 RTMDT INITIAL:
55F RT AFTER 16 EFP: 217 0 F 1800 MATERIAL PROPEPTY 9 SIS AT 3/ T LOCATION 1700 CONTROLLING RA TER IAL: LOWER SHELL COPPER CONTENT: 0.14 wTI, PHOSPHORUS CONTENT: 0.01 4 WT7 1600 R T M INITIAL: 820F_
. RTNDT AFTER 16 EFPY:
163oF 1500 1400 CURVE APPLICABLE FOR HEATUP RATES UP TO 60oF/HR FOR THE SERVICE PERIOD 1300 UP TO 16 EFPY AND CONTAINS MARGINS OF 10oP AND 60 PSiG FOR POSSILE..
INSTRUM'ENT ERROR.
- 12007, 1100
?-TICA L TY I
L 1
-BASED ON INSERVI C HYDROSTATIC TEST 90 TEMPERATURE (3380F FOR THE SERVICE
--71 PERIOD UP TO 16 EFPY 800
- 1.
700 I HEATUP RATES 0F/HR -
6 0 0 o
20 500 60 400 300 Ficure 3.1.3a San Onofre Unit No. 1 Reactor Coolant System 200 Heatup Limitations Applicbie for the Frst 16 EFPY 100
.100 200 300 40 5 0600
-INDICATED TEMPERATURE ('F)
- 7.
2500 H A iL
~P R?~s 240 COUROLLNG Pki"R 1A L INTERMEDIATE SHELL, COPPER CONW 0..
18
.V 2300 R T IAL: 55cF RTNOT AFTER 16 E rPY: 1 14T, 217 OF 2200 I; RV APLCBEFRCODW AE UP 2100 TO1OI/RFRTH EVIEPRO UP TO.
I 60 PSIG FOR POSSIBLE INSTRUm1ENT ERRORS.
- 1900 f...............
1900 1800 17.1 160 0 7-1500
~-
1400 ~
1300-LI.7 1200 000 7 -
-- 7.
7_
9000-,/_
S00 7- --
.70 COOLDOWN RATE 0F/HR
/ f 600-0,,
0 40 300CTE 60-~k~p~(F
400 300 200 ISO 00 80 60 40 0.30% CU BASE, 0.25;1 CU WELD 0.25) CU BASE, 0.20% CU WELD 0.20. CU BASE, 0.15% CU WELD 20 0.15 CU BASE. 0.10% CU WELD 0.10% CU BASE. 0.05% CU WELD 0125 1019 02 FLUENCE (NICM -E
>1 HEV)
Figure 3.1.3c Effect of Fluence and Copper Content on 6RTNDT for Reactor Vessel Steels Exposed to Irradiation at 5500F
020 1.0 x 020 5
SURFACE 2
3/4 THICNESS 1019 5
2 17
'0 1 0
5 10 15 20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)
Figure 3.1.3d Fast Neutron Fluence (E > 1 Mev) as a
- Function of Full-Power Service Life
TABLE 3.1.3.1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
Minimum Average 50 ft-1b/35 mil Upper Shelf Material Cu P
NDTT Temp (F) tTNDT Energy (1t4b)
Component Code No.
Type
%)
%)
(OF)
Long.
Trans.
(OF)
Long Trans.
CI. Hd. Dome W7604 A3028 60(11 112 132 72 72.5 Pel Segment W7605-1 A3028
-10 114 134 74 70.5 Poel Segment W7605-2 A3028
-10 90 110 50 122 Pel Segment W7605-3 A302B
-10 106 1?8 68 86 Pool Segment W7606-4 A302B
-10 120 140 80 74 Peel Segment W7605-5 A3028
-10 26 46
-10 109 Pet Segment W76058 A3028
-10 102 122 62 88 Hd. Flanga W7602 A336 mod 60 1
1b 60 Ves. Flange W7603 A338 mod
[0(a1 (b) 60 Inlet Nozzle W7611-1 A336 mod 60(a)
(b) 60 Inlet Nozzle W7811-2 A336 mod 60[a1
[b) 60 Inlet Nozzle W7611-3 A336 mod 8-(al b) 60 Outlet Nozzle W7610-1 A336 mod 60(11 (b]
60 Outlet Nozzle W7610-2 A336 mod 80 (b) 60 Outlet Nozzle W7810-3 A336 mod 60(a)
(b) 60 Upper Shell W7601-3 A3028 0.15 0.014
-10 48 68 8
98.5 Upper Shell W7601-6 A302B 0.16 0.012
-30 64 84 24 104 Upper Shell W7601-7 A302B 0.15 0.014
-20 52 72 12 95.5
- a. Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2.
- b. Only 10F Charpy V-notch data awilab4a. Conservative estimates for NOTT and RTNDT ere used
TABLE 3.1.3.1 (CONT.)
REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
Minimum Average 50 ft-1b/36 mil Upper Shelf MNDTT Temp (F)
RTNDT Energy (ft-lb)
Component Code No.
Type M%)
M (F)
Long.
Trans.
(a F)
Long.
Tra.s Inte. Shell W7601-1 A3028 0.17 0,013 0
57 12081 go 94 76 lpor, Shell W7601-8 A302B 0.18 0.012 10, 93 1001a]
40 97 79 Inter. Shell W7601-9 A302B 0.18 0.014 0
64 115 55 102 72 Lower Shell W7601-2 A3028 0.17 0.013
-20 74 94 34 97 Lower Shell W7601-4 A302B 0.14 0.014
-10 91 111 51 94 Lower Shell W7601-5 A3028 0.14 0.014 10 122 142 82 87.5 Bot. Hd. Peel W7607 A302B
-20 62 82 22 91 Bot. Hd. Dome W7606 A302B 99 119 60 86e Weld 0.19 0.017 01b) 29ta 0
90 HAZ 0 [b) 1 4 1a 0
101
- a. Actual not estimated
- b. Estimated per NRC Standard Review Plan Branch Technical Position MTES 5-2.