ML20206F098
| ML20206F098 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/10/1986 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206F103 | List: |
| References | |
| NUDOCS 8606240218 | |
| Download: ML20206F098 (12) | |
Text
.
p recuq UNITED STATES k
NUCLEAR REGULATORY COMMISSION w AsmNGTON, D. C. 20655 r,
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. DPR-70 The Nuclear Regulatory Comission (the Comission) has found that:
1.
The application for amendment by Public Service Electric and A.
Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) and supplemented October 16, 1985 and dated October 15, 1984 January 30, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, B.
the provisions of the Act, and the rules and regulations of the Comission; There is reasonable assurance (1) that the activities authorized C.
by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; The issuance of this amendment will not be inimical to the comon D.
defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part E.
51 of the Comission's regulations and all applicable requirements have been satisfied.
i Accordingly, the licenst. is amended by changes to the Technical 2.
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:
8606240210 860610 DR ADOCK 05000272 p
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 75, 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.
R THE NUC A LATORY COMMISSION even arg,
c PWR Project Direc ra e #3 Division of PWR Lic n ing-A, NRR
Attachment:
75 Changes to the Technical Specifications Date of Issuance: June 10, 1986
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ATTACHMENT TO L'ICENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE NO. DPR-70 t
DOCKET NO. 50-272 Revise Appendix A as follows:
Remove Pages Insert Pages 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 g
3/4 4-27a B 3/4 4-6 B 3/4 4-6 8 3/4 4-7 B 3/4 4-7 8 3/4 4-8 B 3/4 4-8 8 3/4 4-9 B 3/4 4-9 8 3/4 4-10 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B 3/4 4-11 4
4 l
MATERIAL PROPERTY BASIS i
UPPER LIMIT OF REG. GUIDE TREND CURVES (FIGURE B3/4 4-2)
COPPER CONTENT
- 0.35 WT%
I PHOSPHORUS CONTENT
- 0.012 WT%
RT INITIAL
- 1/4T 236'F NDT
- 3/4T,107'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERICO UP TO 10 EFPY 3000 lg g
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(51 PSIG and 15.4*F)
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Figure 3.4-2 Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable up to 10 EFPY l
SALEM - UNIT 1 3/4 4-26 Amendment No. 75
MATERIAL P:CFERTY SASIS UPPER Littli 0F REG. GUICE TREND CURVES (FIGURE B 3/4 6-2)
COPPER CONTENT
- 0.35 WTt PHOSPHORUS CONTENT
- 0.012 Wit Ri INITIAL
- 0*F RThAFTER10EFPY : 1/4T, 236*F
- 3/4T. 107'F i
CURVES APPLICABLE FOR C00LOOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY l
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Figure 3.4-3A Salem Unit 1 Reactor Coolant System Cooldown Limitaticns Appli-cable up to 10 EFPY (Excluding Instrument Error Margins)
SAID4 - INC 1 3/4 4-27 Amendment No. 75
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f*ATERIAL P:0PE:lTY BASIS l
UPPER L!ti!T OF REG. GUIDE TPEND CURVES (FIGURE B 3/4 4-2) l COPPER CONTENT
- 0.35 WT l
1 PHOSPHORUS CONTENT
- 0.012 WT:
l RT INITIAL
- 1/47. 236*F NOT
- 3/4T,107'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE SERV!CE PERIOD UP TO 10 EFPY lllllll llllIllllllllll lllllllllll llllii Includes instrument error margins I
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Figure 3.4-3s Salem Unit 1 Reactor Coolant System Cooldown Limitations Appli-cable up to 10 EFPY (Including Instrument Error Margins)
SAI2M - IMIT 1 3/4 4-27a Amendment No. 75 i
6 l
BASES 3/4.4.g PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature end pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles l
used for design purposes are provided in Section 4.1.5 of the FSAR. During startup and shutdown, the ratas 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.
t During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to l
alleviate the tensile stresses induced by the interpal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite O-heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the i
vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Consequently, for the cases in which the outer well of the vessel becomes i
the stress controlling location, each heatup rate of interest must be analyzedonanindfvidualbasis.
The heatup limit curve Figure 3.4-2, is a composita curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves Figure 3.4-3, are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 10 EFPY.
l SALEM - UNIT 1 B 3/4 4 6 Amendment No. 75
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t REACTOR COOLANT SYSTEM 8ASES The reactor vessel materids have been tested to determine their initial RT
- the results of these tests are shown in Table B 3/4.41.
Reactor ophItion and resultant fast neutron (E>l Nev) irradiation will e
cause an increase in the RTgy. Therefore, an adjusted reference taperature, based upon the riuence and copper content of the material in question, can be predicted using Figures 8 3/4.4-1 and 8 3/4.4-2.
The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) include predicted adjustments for this shift in RT at the end of 10 EFPY, as well as adjustments for possible errors in Ne pressure and temperature l
y sensing instruments.
The actual shift in RT of the vessel material will be established periodically during operat15k by removing and evaluating, in accordance with ASTM E185-70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the Since the neutron spectra at the irradiation samples and core area.
vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.
The hentup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different fromtheca150IatedART for the equivalent capsule radiation aposure.
NOT The pressure-taperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure empliance with the minimum taperature require-ments of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation ' surveillance specimens and the frequencies for removing and testing these specimens are in accordance with the requirements of Appendix H to ' 0 CFR Part 50.
The41mitations imposed on pressurizer heatup and cocidown and spray water taperature 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.
l SALEM - UNIT 1 8 3/4 4-7 Amendment No. 75
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A 111 ll Im l ' m I %lll$ 7.,, ' ], ' l l 4 l ' it o .I ll i towen uld,1ll y g!' E!'l if %e - e.oTi 1 ) m i. U l ti % Cu = 0.08 l l i l 3 - y i i I ~ ~ % P = 0 00,8 I I .I , i 3 a r.i -l E i i ( A-1 l 22:-- 2X1oli 4 S S to 8 i i I 2 4 8 8 lo 8 l 2 4 s FLUENCE, en/en 2 (E > 1MeV) h* 8 e Figure B3/4.4-2 Effect of Fluence and Copper and Phosphorus Contents on ART g NOT for Reactor vessel steels g & Weld Metal (Based on Capsule Y results) G Shell Plate B2402-3 (Based on Capsule Y results)
M r-l j TABLE B 3/4.4-1 SALEM UNIT 1 REACTOR VE5SEL 100Glett55 DATA (UNIRRA01ATED) I 50 ft Ib 35-Mil y RT Average Shelf Inergy Material Cu P IIST Teep NOT Ince It#D e Campenent Heat No. Gode No. Type (1) (1) (*F) (*F) (*r) (ft Ib) (ftIb) C1 Nd Dome A0610 82407-1 A5338. C1.1 0.20 0.011 -30 99* 39 110 C1 Nd Segment C1544 82406-1 A5330. C1.1 0.13 0.010 -20 89= 29 125 C1 Nd Segment C1544 82406-2 A5338. C1.1 0.16 0.012 -30 85* 25 122-C1 hd Segment 85852 82406-3 A5338. C1.1 0.10 0.009 -50 66* 6 132 Cl Nd Flange 123P409 82811 A508. C1.2 0.010 28* 22* 28 199 Vessel Flange SP1191 82410 A500. C1.2 0.009 60* 0* 50 145 N i Inlet Nearle 123P403 82408-1 A508. Cl.2 0.010 50* 43* 50 144 Inlet Ilozzle 125P544 32408-2 A508. C1.2 0.011 46* 26* 46 157 Inlet IIerale 123P403 82408-3 A508. C1.2 0.010 47* 37* 47 161 Inlet florale 125P544 B2408-4 A508. C1.2 0.010 9* 17* 9 167 Outlet Nozzle ZT2550 82409-1 A508. C1.2 0.010 60* 95* 60 75 Outlet IIstrie ZT2550 82409-2 A500. C1.2 0.011 60* 95* 60 ' 78 R Outlet IIsrzle ZT2585 B2409-3 A508. C1.2 0.013 60* 10* 60 121 Outlet Nozzle ZT2585 B2409-4 A508. C1.2 0.012 60* 13* 60 126 i t W r Shell A0497 B2401-1 A5338. C1.1 0.22 0.012 -30 87* 27 114 g upper Shell A0495 82401-2 A5338. C1.1 0.19 0.011 0 80* 20 122 upper Shell A0512 B2401-3 A5338. Cl.1 0.24 0.011 -10 114e 34 96 later Shell C1354 82402-1 A5338. Cl.1 0.24 0.010 -30 105 45 73.0 97 Inter Shell Cl354 B2402-2 A5338. C1.1 0.24 0.010 -30 55 -5 91.5 112 Inter Shell C1397 82402-3 A5338. C1.I 0.22 0.011 -40 57 -3 104.0 127 teuer Shell C1356 82403-1 A5335. C1.1 0.19 0.011 -40 70 10 99.0 143 g Leuer Shell C1356 S2403-2 A5330. Cl.1 0.19 0.012 -70 86 26 94.0 128 5 tener Shell C1356 32403-3 A5338. C1.1 0.19 0.010 -40 90 30 102.0 131 "g. est Nd Segeant A0705 82404-1 A5338. Cl.1 0.10 0.009 10 48* 10 120 g tot Nd Segment A0705 B2404-2 A5338. C1.1 0.11 0.010 -50 60* O 132 3 tot Nd Segment A0705 82404-3 A5338. C1.1 0.12 0.000 10 47* 10 126 i Sat Md Osse A0705 82405-1 A5338. C1.1 0.15 0.010 -20 57* -3 106 2 % illa m 0.16 0.019 0* - 38 *
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1888 - Normal to Major Working Direction 1810 - Major Working Direction - Estimated per IIItC Standard llevleu Plan Branch Technical Position MTE8 5-2 j Actual transverse data obtained from surveillance program (from minimum data points). I
REACTOR COOLANT SYSTEM BASES The OPERASILITY of two POPSs or an RCS vent opening of greater than 3.14 souare inches ensures that the RCS will be protected from press fe transients which could exceed the limits of Appendix G to 10 CFR Pa-- 50 when one or more the RCS cold legs are less than or equal to 312'F. Either POPS has adequate relieving capability to pretect the RCS ft:- overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the stear generator less than or eaual to 50*F above the RCS cold leg temperatures, or (2) the start of a safety injection pump and its injection into a wate solid RCS. I!
- p. 1 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 comoonents
= ensure that the structural integrity of these components will be reintain' at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance ll with Section XI of the ASME Boiler and Pressure Vessel Code. SALEM - UNIT 1 B 3/4 4-11 Amendment No. 75 /}}