ML19317G782
| ML19317G782 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/16/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G777 | List: |
| References | |
| TAC-08884, TAC-8884, NUDOCS 8004010608 | |
| Download: ML19317G782 (9) | |
Text
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/jamaag**g UNITED STATES 4' '
NUCLEAR REGULATORY COMMISslON
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j VlASHINGTON, D, C. 20666 e,
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%,.....f SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amenchnent No. 22 License No. DPR-54 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Sacramento Municipal Utility District (the licensee) dated August 2,1978, as revised August 11, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regCattons of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted withot.t endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to i.he comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
8004010h07 I
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2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license
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amendment, and paragraph 2.C.(2) of Facility Operating License i
No. DPR-54 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 22, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
1 FOR THE NUCLEAR REGULA COP 91ISSION yh l(82.
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: August 16, 1978 e
m ATTACHMENT TO LICENSE AMENDMENT NO. 22 FACILITY OPERATING LICENSE NO. OPR-54 DOCKET NO. 50-312 Revise Appendix A as follows:
Remove Pages, Insert Pages 3-3 3-3 3-3a 3-3a 3-4 3-4 Figure 3.1.2-1 Figure 3.1.2-1 Figure 3.1.2-2 Figure 3.1.2-2 Figure 3.1.2-3 3-5 The changes on the revised pages are shown by marginal lines.
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N RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.2.
PRESSURIZATION, HEATUP, AND COOLDOWN LIMITATIONS Spacification 3.1.2.1 Inservice Leak and Hydrostatic Tests:
Pressure temperature limits for the first five EFP years of inservice leak and hydrostatic tes ts are given in Figure 3.1.2-3.
Heatup and cooldown rates shall be restricted according to the rates specified in Figure 3.1.2-3.
3.1.2.2 Heatup Cooldown:
For the first five EFP years of power operation the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in.accordance with Figure 3.1.2-1
.tud Figure 3.1.2-2 respectively. Heacup and cooldown rates shall not exceed the rates stated on the associated figure.
3.1. 2. 3 The secondary side of the steam generator shall not be pressurized
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above 200 psig if the temperature of the steam generator shell is below 1300F.
3.1.2.4 The pressurizer heatup and cooldown' rates shall not exceed 1000F in any 1-hour period.
3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 4100F, 3.1.2.6 Prior to exceeding five effective full power years of operation, Figures 3.1.2-1, -2 and -3 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.B.
The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff a
review in accordance with Specification 3,1,2.7, 3.1.2.7 The updated proposed technical specifications referred to in 3.1,2.6 shall be submitted for, NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C.
Amendment No. 22 3_3
,N RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
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Bases 23 pressure-camperature limits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to 10 CFR50 i
it and with the thermal and loading cycles used for design purposes.
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- l Ths limitations prevent non-ductile failure during normal operation, including ll anticipated operational occurrences and system hydrostatic test. The limits l;
also prevent exceeding stress limits during cyclic operation. The loading
- l conditions of interest include
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1.
Normal heacup 2.
Normal cooldown l
3.
Inservice leak and hydrostatic test The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. The closure head ragion, reactor vessel outlet nozzles and the beltline region have been identified to be the only regions of the reactor vessel, and consequently of the reactor coolant pressure boundary, that determine the pressure-tempera-ture limitations concerning non-ductile f ailure.
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'l the closure head region is significantly stressed at relatively low ctmperatures (due to mechanical loads resulting from bolt pre-load). After temperature of the beltline 5 EFPY of neutron irradiation exposure, the RINM region materials will be high enough so that the beltline region of the r3 actor vessel will control much of the pressure-teenerature limitations of the reactor coolant pressure boundary. For the. service period for which the limit curves are establisned, the maximum allowable pressure as a func-tion of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and belt-line region.
Amendment No.
, 22' 33,
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RANCHO SECO UNIT 1 TECHNICAL SPEf*IFICATIONS Limiting Conditions for Operation The maximum allowable pressure is taken to be the icwest pressure of the three esiculated pressures. The pressure limit is adjusted for the pressure differencial batveen the point of system pressure measurement and the limiting component for all reactor coolant pump conbinations. The limit curves were prepared based upon the most limiting adjusted reference temperature of all the beltline region matarius at the end of the fif th effective full power year.
Ihn actual shif t in RT of the beltline region material will be established p3riodically during ope $ tion by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irrr.diation surveillance specimens installed near the inside wall of this or a.similar reactor vessel in the core area.
Because 'the neutron energy spectra at the specimen location and at the vessel inner wall location are essentially the same, the measured transition shif t for a sample can be applied with confidence to the adjacent section of the reactor determined from vacsel. The limit curves must be recalculated when the 6RTNDT the surveillance capsule is dif forent from the calculated ARIgg for the equiva-Inne capsule radiation exposure.
The unirradiated impact properties of the beltline region materials, required j
by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available. The adjus ted reference temperatures are calculated by adding the radiation-induced ARTNDI and the unirradiated RINDr*
are caleclated using the respective neutron fluence and The predicted ARTgg copper and phosphorus contents in accordance with Reg. Guide 1.99.
' Die assumed RT of the closure head region is 600F and the outlet nozzle steel forgings is 60k The limitations imposed on pressurizer heacup and cooldown and spray water ccuperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME code requirements.
1 Amendment No. f,
, pp, 3_4
l RANCIO SECO 1 REACTOR COOLANT SYSTEH PRESSURE-TEMPERATURE LIMITS FOR HEATUP FOR THE FIRST 5 EFPY MAXDt#1 HEATUP RATES E
TEW MAX RATE 2500 7
0 70*F-100* F 5*F/HR h
N 100*F-135*F 25*F/HR E
Restricted Region 2000 135*F-185'r 50*F/HR e
3E 185*F-550*F.
75*F/HR S
Od POINT PRESS TEW
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1500 A
336 70 0
8 336 122 2
C 310 175 Permissible D
525 295 Operating E
2250 458 N"9 ""
1000 8
5.
3u MARGINS OF 25 PSIG 10*F ARE 7
500 D
INCLUDED FOR POSSIBLE INSTRUE NTA-y TION ERROR E
A B
C
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l 100 200 300 400 500 Indicated Reactor Coolant Temperature. Tc. *F Amendment No. 22' FIGURE 3.1.2-1
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RANCHO SECO 1 REACTOR COOLANT SYSTEM PRESSURE-TEiFERATURE LlHITS FOR C00LOOWN FOR THE FIRST 5 EfPY g
2500 E
G b
C00LOOWN RATE D
5 TEW MX C.O. RATE g
2000 j
550-280 100*/HR 5
280-150 50*/HR i
150-25*/HR Restricted Region tj 1500 he s*
POINT PRESS TEW pe t g
Region A
150 80*
e F
E B
275 125*
1000 g
C 225 163*
O O'
0 315 250*
s.
b E
E 525 300*
s oc 500 F
1250 362*
3 0
G 2250 407*
g l
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5 A
100 200 300 400 500 MARGINS OF 25 PSIG and 10*F ARE INCLUDED FOR POSSIBLE INSTRUMEN-Indicated Reactor Coolant Temperature. T, F TATION ERROR c
FIGURE 3.1.2-2 Amendment No. ] 8 l
RANCHO SECO 1 INSERVICE LEAK AND HYDROSTATIC TEST (5 EFPV) HEATUP AND C00LDOWN 2500 -
y HAX HEATUP RATE i 50*F/HR HAX C00LDOWN RATES:
G TEMP C.D. HAX RATES
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550-280 100*F/HR 2000 Restricted Region 280-150 50*F/HR 150-25'F/llR POINT PRESS TEMP 1500 Permissible A
230 75 Operating b
8 380 132 1
E F
C 370 144 jogo 0
330 165 E
575 275 E
F 950 322 500 G
2250 419 C
8 t
H 2500 426 A
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I HARGINS OF 25 PSIG and 10*F ARE l
100 200
- 300, 400 500 INCLUDED FOR POSSIBLE INSTRUMEN-
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TATION ERROR Temperature. *F Amendment No.
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FIGURE 3.1.2-3 N'
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