ML20213A663
| ML20213A663 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/14/1987 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213A664 | List: |
| References | |
| NUDOCS 8704280151 | |
| Download: ML20213A663 (6) | |
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NUCLEAR REGULATORY COMMISSION j
WASHING TON. D. C. 20666
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SOUTH CAROLINA ELECTRIC A GAS COMPANY SOUTH CAROLINA PU9LIC SERVICF AUTHORITY
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DOCKET NO. 50-395 VIRGIL C. StMMER NUCLEAR STATION UNIT NO. 1 AMENDMENT TO FACILITY OPEDATING LICENSE Amendment No. 65 License No. NPF-12 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by South Carolina Electric & Gas Company and South Carolina Public Service Authority (the licensees) dated January 8,1987, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comissioni C.
There is reasonable assurance (il that the activities authorized 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 regulationst D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the publict and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable recuirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.f 2) of Facility Operating License No. NPF-12 is hereby amended to read as follows:
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l (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 65
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This amendment is effective as of its date of issuance, and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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t Elinor G. Adensam, Director Project Directorate Il-1 Division of Reactor Projects-!/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 14, 1987 i
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ATTACWENT TO LICENSF AMENDMFNT 4
i APENDMENT P0. 65 TO FACILITY OPEPATING LICENSE NO NPF-12 DOCKET NO. 50-395 Replace the followin the enclosed pages. g pages of the Appendix "A" Technical Specifications with The revised pages are identified by amendment number and contain vertical lines jndicating the areas of change. Corresponding overleaf l
pages are also provided to maintain document completeness.
Remove Pages Insert Pages 2-7 2-?
R2-5 B2-5 l
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h TABLE 2.2-1 (continued)
A REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SE Functional Unit Total Allowance (TA)
Safety injection Input Z
S O
18.
Trip Setpoint from ESF NA Allowable Value NA NA NA NA
- 19. Reactor Trip System Interlocks A.
Intermediate Range NA Neutron Flux, P-6 NA NA
>7.5 x 10.s%
Low Power Reactor Trips Indication
>4.5 x 10 8%
B.
i Indication Block, P-7 a.
P-10 input
- 7. 5 4.56 0
m b.
P-13 input 110% of RTP I
4 7.5
$12.2% of RTP 4.56 0
<10E turbine Impulse pressure
<12.2% of turbine C.
Power Range Neutron equivalent impulse pressure Flux P-8
- 7. 5 equivalent 4.56 0
<38% of RTP
<40.2% of RTP D.
Low Setpoint Power
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7.5 Range Neutron Flux, P-10 4.56 0
Turbine Impulse Chamber 7.5 Pressure, P-13 4.56 0
<10% turbine Impulse pressure
<12.2% turbine y
F.
Power Range Neutron equivaient jiressure equivalent Flux, P-9 7.5
- 4. %
0
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- 20. Reactor Trip Breakers MA NA NA NA y
21.
Automatic Actuation logic NA NA NA NA NA NA O
RIP = RATED THERMAL POWER 1
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_ LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Nuclear Flux (Continued) i uncontrolled rod cluster control assembly bank withdrawal f condition.
These trips provide redundant critical of the Power Range, Neutron Flux channels. protection to the low setp initiate a reactor trip at about 10+5 The Source Range channels will when P-6 becomes active.The purpose of the P-6 setpointcounts per second lower end of the intermediate range scale, is to give the, which time to actuate the source range reactor trip block operators sufficient THERMAL POWER unless manually blocked
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I Overtemperature AT ve.
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The Overtemperature delta T for all combinations of pressure, trip provides core protection to prevent transit delays from the core to the temperature power, coolant temperature, and axial power spect to piping and pressure is within the range between the Pressurizer h a
trips.
The setpoint is automatically varied with 1) coolant te correct for temperature induced changes in density and hea n
ow pressure temperature detectorsand includes dynamic compensation erature to y of water
- 2) pressurizer pressure e to the loop With normal axial powe,r distribution, this reac, tor triand 3) axial the core safety limit as shown in Figure 21-1 p limit is always below than design, as indicated by the difference between topIf axia nuclear detectors, the reactor trip is automatically reduce notations in Table 2.2-1.
t r ng to the Overpower AT fuel melting and less than 1 percent cladding st egrity (e.g., no overpower conditions, limits the required range for Overtemp er all possible protection, and provides a backup to the High Neutron Fl is automatically varied with 1) coolant temperature to corre ure delta T ux trip.
The setpoint induced changes in density and heat capacity of water of temperature for dynamic compensation for piping delays f or temperature
, and 2) rate of chance loop temperature detectors to ensure that the allowable h (Kw/ft is not exceeded.
m the core to the the con) sequences of various size steam breaks as eat generation rate Core Response to Excessive Secondary Steam Break " ported in WCAP Pressurizer Pressure In each of the pressure channels thus limiting the pressure range in which reactor low setpoint trip protects against low pressure which could l pressure trip on is permitted.
The tripping the reactor in the event of a loss of reactor c ead to DNB b oolant pressure. y SUMMER - UNIT 1 B 2-5 Amendment No. 65 o
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.2 LIMITING SAFETY SYSTEM SETTINGS i
i BASES Pressurizer Pressure (Continued)
On decreasing power the low setpoint trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED l
and on increasing power., automatically rejnstated by P-7..
J The high setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system l
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overpressure, Pressurizer Water Level l
The pressurizer high water level trip is provided.o prevent water r through the pressurizer safety valves.
r water level trip is automatically blocked by P-7 (a pov ir level of approximately 10 percent of RATED THERMAL POWER with a turbine impuls4 chamber press approximately 10 percent of full equivalent); and on ir;reasing power, l
l automatically reinstated by P-7.
Loss of Flow The Loss of Flow trips provide core protection to prevent DN6 by mitigating
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the consequences of a loss of flow resulting from the loss of one or more i
reactor coolant pumps.
j On increasing power above P 7 (a power level of approximately 10 percent of RATED THERMAL POWER or a turbine impulse chamber pressure at approximatel i
10 percent of full power equivalent), an automatic reactor trip will occur if l
the flow in more than one loop drops below 90% of nominal full loop flow.
Above P-8 (a power level of approximately 38 percent of RATED THERMAL POWE an automatic reactor trip will occur if the flow in any single loop drops Conversely on decreasing power below 90 percent of nominal full loop flow.
between P-8 and the P-7 an automatic reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically j
blocked.
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Steam Generator Water Level The steam generator water level low-low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mis The specified setpoint provides l
resulting from loss of normal feedwater.
allowances for starting delays of the auxiliary feedwater system.
1 Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The steam /feedwater flow mismatch in coincidence with a steam generator i
Iow water level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional cappbility of the specified trip settings and thereby enhance the overall reliability of the Rea i
i Protection System.
The Steam /Feedwater Flow Mismatch portion of this trip is Low-Low trip.
activated when the steam flow exceeds the feedwater flow by greater than or The Steam Generator Low Water level portion i'
equal to 1,63 x 10' lbs/ hour.of the trip is activated when the water level d l
SUMMER - UNIT 1 8 2-6
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