ML19319E184
| ML19319E184 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/13/1976 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19319E176 | List: |
| References | |
| NUDOCS 8003310700 | |
| Download: ML19319E184 (8) | |
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un:Td3 STATES
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- NUCLEAR REGULATORY CC.'. MISSION -
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WASHINGTON, D. C. 20655 xy..st)
SACRA!TfTO MU'iICIPAL UTILITY DISTRICT DOCKET No. 50-312-RANCHO SECO NUCLEAR GENERATING STATION
-AMENDMENT TO FACILITY OPERATING LICEKSE Amendment No. 7 License No. DPR-54 1.
The Nucicar Regulatory _. Comission (the Commission) hn found that:
A.
The application for acendcant by the Sacramento Municipal Utility District (the licensee)~ dated April 8, 1976, as supplemented April 28, 1976, and May 21, 1976, 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 ferth in 10 CFR Chapter I;
'B.
The facility will operate in conformity with the application,
'the provisions of the'Act, and the~ rules and regulations of the Coc=ission; C.
There is reasonable assurance (1) that the activities authori::ed by this amendment can be conducted without endangering the health and safety of the public, and (ii) that.such activities will be Jco1 ducted f.a ccmpliance with the-Commission's regulations; D.
The issuance of this amendment will not be inimical to the cocmon 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 51 of the Comsission's regulations and all applicable requirements have been satisfied.
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Accordingly,.the licenseIis amended by a change in the' Technical
-Specifications as indicated in the attachment to this license i
. amer dment.
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This license amendment is effective as of the date oi 'ts issuance.
FOR TIIE NUCLEAR REGULATORY CO:OilSSION Q M bl Robert W. Reid, Chief Operating Reactors Branch #4 Division of' Operating Reactors Attach:nent:
Changes to the -.
. Technical Specifications Date of Issuance: Au, st 13,1976 we 4
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- ATTAC11 MENT TO LICENSE AMENDMENT NO. 7 FACILITY OPERATING LICENSE NO. DPR-54.
DOCKET No.~ 50-312-
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Revico. Appendix A as follows:
Remove Pages Insert Pages:
3-3a 3-3a 3-4
~3-4
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4-7b-4-7b 4-8 4-8 4-12 4-12 Changes on the revised pages are shown by_a marginal 11ne. Pages
'4-7b and 4-12'are unchanged and are included,for convenience only.
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PANC110 SECO UN!T *.
.TEC dCA!.: SPECIFICATIONS
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Limiting' Conditions for Operation -
, -- 3.1'.'2. 5. The pressusizer he,stup and cooldown rates shall not exceed 100 F/hr.
.The spray shall not be used if t.he tc=;c'rature difference between
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the pressurizer and the spray fluid.ic greater than 410 F..
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-3.1.2.6_ L',ithin two years of power operation, figures 3.1.2-1 and 3.1.2-2 shall be. updated in' accordance with appropriate criteria accepted i
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-by the AEC.
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~.All, reactor coolant' system components are designed to withstand the effcces
.cf cyclic loads ' due. to system temperature and pressure changes. (1) TIese cyclic N
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~3-3a Amendment.No. 7
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RANCHO SECO UNIT 1 5Q
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- TEClfNICAL OPECIFICATIONS J k j
_;3 Limiting conditions for Operation r
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loads arc' introduced by unic load transients, reactor'erips, and unit heatup
- and cooldova operations-The number of thermal and lording cycles used for
~ design purposes nra'shown in table 4.1-1 of the FSAR. The maximta unit heatup
- und cooldown rate of 100 F per hour satisfics stress limits for cyclic opera-
, tion.('d Tha 200 poig pressure linit for thu secondary side of L'nc steam genc'rator at a temperature less than 130 F satisfies -stress levels for tem-peraturcs below the DTT.(3) The reactor vessel place material and r ids have b,cen tested to verify conformity to specified requirements cnd a :. r
.m NDTT value of 10 F has been determined based on Charpy V-notch tests Ti' naximum NDTT value obtained-for the steam generator shcIl material and ucide was 70 F.
Figures 3.L 2-1 and.3.1.2-2 contain the limiting reactor coolant system pressure-temperature relationship for operation at.DIT(0) and below to, assure that stress icvels are low enough'to preclude brittic fracture. These stress levels and their bases are defined in paragraph 4.3.3 of the FSAR.
As a result of fact-neutron irradiation in the region of the core, there will be an increase in the NDTI with accumulated nuclear operation. The predicted maxihua NDTT increase for the 40-year exposure is shown on figurc 4.3-1 of the FSAR.(4).The accu 1. shift in NDIT will be determined periodically during plant operation by testing'of irradiated vessel. material samples located in this reactor vessel. (5) The results of the irradiated sa=ple testing will be evalu-ated and compared to the design curve (figure 4.3-2 of FSAR) being used to prc-dict the increase in transition temperature.
The design value for fast neutron (E >.1 McV) exposure of the reactor vessel is 3.0 x.1010 n/cm2ccc at 2,772 MI t. rated nower and an integrated c::posure of for 40 years operacion.(d The calculated maximum values 3.0 x 1019 n/cm2 are 2.4 x 1010 n/cm2see and 2 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load.
Figure 3.1.2-1 is based on the design value which is considerably higher then the calculated value. The DTT value for
' figure 3.1.2-l' is based on the projected UDTT at the end of the first two.
years of operation. During these two years,.the energy output has been con-servatively estimated ~to be 1.8 x 106 thermal megawatt days, which is equiva-lent.co 655 days at 2,772 MWt core power. The projceted fast neutron exposere 13 n/ct2 of the reactgr vessel for the two years 'is 1.7 x 10 whi~ch is based on the 1.8 x'10 thermal megawatt days ~and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodically during plant opera-tion by testing vessel material samples which are irradiated by securing then l
.ne'ar the inside vall^of the vessel in the core area. To compensate for the increases.in the NDTT caused by irradiation, the li=its on the pressure-temperature relationship are periodically changed to stay within'the established stress limits during heatup and cooldown.
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-.The NDTT shift.and the maanitude of the thermal a..d prcosure stresses are sen-sitive co-integrated reactor power and not to'instantanceus novar icvel.
Fig'ures 3.1.2-1 and L3.1.2-2 are' applicable to reactor core ther:al. ratings up to 2,772 Wt. -
t AmebdmentNo.7.
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TABLE 4.1-1 (Continued).
b INSTRUMENT' SURVEILLANCE REQUIREMENTS
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-Channel' Description
- Check.
Test.
. Calibrate Remarks
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R'eactor. Building drain NA.
NA-R b..
accumulation tank level'
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43.' 'Incore neutron detectors M(1)
NA NA (1) Check functioning, includ-ing functioning of computer readout and/or recorder y
j readout c)
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- 44. ' Process and area radia-W H
Q tion monitoring systems p @.
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Emergency plaat radiation M(1)
NA
'R (1). Battery check co
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instruments J
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Environmental air
'H(1)
NA R
(1) Check functioning y2 monitors
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NA R
(1) Battery check d
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5' accelerometer a
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S = Each shift M = Honthly P '= Prior to each.startup if not donc previous week. '
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. Quart'erly R = Once during the refueling interval
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D = Daily E
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RANCHO SECO UNIT NO. 1 TECHNICAL SPECIFICATIONS t-Surveillance Standards TABLE 4.1-2 MINIMUM EQUIPMENT. TEST FREQUENCY Item Test Frequency 1.
Control rods Rod drop times of.
Each refueling shutdown all full length' rods 2.
Control rod Movement of each Every two weeks movement rod 3.
Pressurizer code.
Setpoint 1 each refueling interval safety valves c.
4.
Main Steam safety Setpoint 2 per steam generator valves each refueling interval 5.
Refueling system-Functional Each refueling interval interlocks prior to handling fuel.
6.
Turbine steam stop Movement of each Monthly
. valves valve 7.
Reactor Coolant Leakage Calculated inventory weekly system Leakage check daily.
8.
Charcoal and high Charcoal and HEPA Each refueling interval and efficiency filters filter for iodine at any time work on filters i
and particulate could alter their integrity 1
I removal efficien-cies. 00P test
'on HEPA-filters.
Freon test on i
charcoal filter units.
9.
Fire pumps and Functional Monthly power supplies
- 10. Reactor Building Functional Each refueling interval
~ isolation trip-
- 11.. Spent-fuel Functional Each refueling interval cooling system prior to fuel handling.
12".. Turbine Overspeed Calibration Each refueling interval Trips.
13.' internals Vent ~
Manual Actuation,I Each refueling interval
.spection,{sualin-
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Valves Remote V and verify
' that valve not stuck open.
= 1. Verify that the valve begins to open from.the fully closed position with a force equivalent to < (0.15)-psid and is fully open with a force equivalent surTac(0.30) psid. 27 Check-visually accessible surfaces to evaluate obser_ved -
to <
e, irregularities.
Amendment No. 7 4
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J RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.2-1 CAPSULE ASSEM3LY WITdDRAWAL SCHEDULE
' Surveillance. spec!:2cas will be-withdrawn at appro::imately 170
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4 ef fective' full poucr days and reinserted at the end of the first i
fuel cycle to allow modifications to be incorporated into the -
specimen holder configuration. Thereaf ter, capsulcs will be with-
-,. drawn in accordance with-the following schedule:
First capsule--
At the time when predicted shif t of Cy adjusted fracture energy curve is approxi-mately 50 F. or at one-fourth service lifs, whichever is earlier.
Second and third -
-At apprbximately one-third and two-thirds -
capsules of the time interval between first and fourth capsule withdrawal.
Three-fourths of service life.
. Fourth capsule---
. Fif th capsule-----------Standby.
9 4-12 Amendment No. 7 e
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