ML20078P386
| ML20078P386 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/19/1983 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078P389 | List: |
| References | |
| NUDOCS 8311080026 | |
| Download: ML20078P386 (9) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION p
WASHINGTON, D. C. 20555
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VIRGINIA ELECTRIC AND POWER COMPANY' DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 3P License No. NPF-7 i
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Virginia Electric and Power Company (the licensee) dated June 8, 1982 and May 3, 1983 comply 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; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) 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 regulations; D.
The issuance of this amendment will not be inimical to the common 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.
8311000026 831019 PDR ADOCK 05000339 P
PDR l
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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.(2) of Facility Operating License No.
NPF-7 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 32
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is e'ffective within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION O
W A
<g/ James R. Miller, Chief 1
Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 19, 1983 4
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT N0. 32 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET N0. 50-339 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
I Pages 2-8 i
2-9 2-10 3/4 2-16 1
4 4
-,e
-.,,,.,y-
2 TABLE 2.2-1 (Continued)
ES l
52 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ji FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 13. Steam Generator Water
> 18% of narrow range instrument
> 17% of narrow range instrument Ei Level--Low-Low span-each steam generator span-each steam generator H
- 14. Steam /Feedwater Flow
< 40% of full steam flow at
< 42.5% of full steam flow at n3 Mismatch and Low Steam NATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level
> 25% of narrow range instru-3 24% of narrow range instru-ment span--each steam generator ment span--each steam generator
- 15. Undervoltage-Reactor
> 2905 volts-each bus
> 2870 volts-each bus Coolant Pump Busses
- 16. Underfrequency-Reactor
> 56.1 Hz - each bus
> 56.0 Hz - each bus Coolant Pump Busses n,
- 17. Turbine Trip A.
Low Trip System
> 45 psig
> 40 psig Fressure B.
Turbine Stop Valve
> 1% open
> 0% open Closure
- 18. Safety Injection Input Not Applicable Not Applicable from ESF 1
- 19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip
4 is TABLE 2.2-1 (Continued) 4 2
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
JE NOTATION NOTE 1:
Overtemperature AT < AT, [K -K2 3
}~ 1(O }
1 (T-T')+K ( ~
j
-4 1+T S 2,
n3 where:
AT, Indicated AT at RATED THERMAL POWER
=
Average temperature, "F T
=
Indicated T at RATED THERMAL POWER < 582.8"F I
T'
=
avg l
Pressurizer pressure, psig P
=
h[
P' 2235 psig (indicated RCS nominal operating pressure)
=
1+t S j
The function generated by the lead-lag controller for T,yg dynamic compensation
=
- j. 3 E
Time constants utilized in the lead-lag controller for T t = 25 secs, t
t =
1 2
- U t = 4 secs.
2 Laplace transform operator (sec~1) f[
S
=
e a
if O
k!
5 TABLE 2.2-1 (Continued)
E*
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SET?OINTS NOTATION (Continued) i g
Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops (no loops isolated)*
(1 loop isolated)*
-a
(
)
(
)
K K
=.l.141 K
=
=
g g
g
(
)
(
)
K 0.0128 K
K
~
=
2 2
2
(
)
(
)
K 0.000608 K
K
~
=
=
3 3
3 1
and f (AI) is a function of the indicated difference between top and bottom detectors g
of the power-range nuclear ion chambers; with gaint to be selected based on measured instrument response during plant startup tests such that:
1 (i) for q
- gb between - 35 percent and + 7 percent, f (AI)
=0
'f (where q and q arepercentRATEDTHERMALPOWERintbetopandbottom g
b l
halves of the core respectively, and q
+ q is total H E N POWER in b
i percent of RATED THERMAL POWER).
(ii) for each percent that the magnitude of (q
- q ) exceeds - 35 percent, b
4 the AT trip setpoint shall be automaticalky reduced by 1.58 percent of its value at RATED THERMAL POWER.
(iii) for each percent that the magnitude of (q
- q ) exceeds + 7 percent, b
i the AT trip setpoint shall be automatical$y reduced by 1.24 percent of 3
its value at RATED THERMAL POWER.
N
.E ea 1
N:
{
- Values dependent on NRC approval of ECCS evaluation for these operating conditions.
e---
TABLE 2.2-1 (Continued) g d
2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS N
NOTATION (Continued) g s
5 3
T-K6 (T-P)-f (O )3 Note 2:
-Overpower AT 1 AT,[K -K5 2
4 1+t 53, where:
AT, Indicated AT at RATED THERMAL POWER
=
Average temperature, F
T
=
Indicated T,yg at RATED THERMAL POWER $582.8 F, l
0 T"
=
1.088 K
=
7 4
0.02/*F for increasing average temperature K
=
5 0 for decreasing average temperatures K
=
5 0.00119 for T > T";
K I
K
=
6 6
I 3"
3 The function generated by the rate lag controller for T,yg i
=
g j
3 3
dynamic compensation g
m Time constant utilized in the rate lag controller for T,yg 5
T
=
3 T = 10 secs.
g 3
S'=
Laplace transform operator (sec~ )
f (AI) =
0 for all AI 2
Note 3:
The, channel's maximum trip point shall not exceed its computed trip point by more than 2 percent span.
O
POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:
Reactor Coolant System T,yg a.
b.
Pressurizer Pressure c.
Reactor Coolant System Total flow Rate APPLICABILITY: MODE 1 ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than S% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l 4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by maasurement at least once per 18 months.
l I
l l
l NORTH ANNA - UNIT 2 3/4 2-15
TABLE 3.2-1 o
ll DNB PARAMETERS is SE LIMITS
[]
2 Loops In Operation **
2 Loops In Operation **
- q 3 Loops In
& Loop Stop
& Isolated Loop Operation Valves Open Stop Valves Closed PARAMETER Reactor Coolant System T,yg
<587*F l
Pressurizer Pressure
>2205 psig*
>278,400 gpm Total Flow Rate
~
kR "f
as
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a 27 THERMAL POWER step in excess of 10% RATED THERMAL POWER.
E it
- Values dependent on NRC approval of ECCS evaluation for these conditions, En 8
.