ML20087P409
| ML20087P409 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/13/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087P410 | List: |
| References | |
| NUDOCS 8404060406 | |
| Download: ML20087P409 (13) | |
Text
.
k' UNITED STATES 8"
NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D. C. 20555
'g p
VIRGINIA ELECTRIC AND POWER COMPANY l
OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. NPF-4 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Virginia Electric and Power Company (the licensee) dated December 30, 1982 (as supplemented April 25, July 6, 1982, July 11, 1983) and September 29, 1983, complies with the standards and requirement.s 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 conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; 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 activitier. will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements have been satisfied.
l 8404060406 840313 i
PDR ADOCK 05000338 P
_-- O
. 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF 4 is hereby amended to read as follows:
(2) Technical Specifications.
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 54, are he aeby incorporated in the license. The licensee shall opera *o :he facility in accordance with the Technical Specificatbes.
3.
This license amendment is effective within 30 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
%f James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 13, 1984
P ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by Amemdment number and contain vertical lines indicat..ig the area of change.
The corresponding overleaf pages are included to maintain document complete-ness.
Pages 2-2 2-6 2-8 2-9 2-10 2-15
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tlimits shown in Figures 2.1-1 shall not exceed the for 2 loop operation.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in H0T STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded ~ 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.
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P103 304 0
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1 NORTH ANNA - UNIT 1 2-z Amendment No. A6, 54 m
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
4 NORTH ANNA - UNIT 1 2-5
TABLE 2.2-1 g
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E$
1.
Manual Reactor Trip Not Applicable Not Applicable I
2.
Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Setpoint - < 26% of RATED ei'i THERMAL POWER -
THERMAL POWER -
.. y High Setpoint - < 109% of RATED High Setpoint - 1 110% of RATED THERMAL POWER THERMAL POWER
- 3.
Power Range, Neutron Flux,
< 5% of RATED THERMAL POWER with
< 5.5% of RATED THERMAL POWER High Positive Rate a time constant > 2 seccnds with a time constant > 2 seconds 4.
Power Range, Neutron Flux, 1 5% of RATED THERMAL POWER with
< 5.5% of RATED THERMAL POWER High Hegative Rate a time constant > 2 seconds
'?
with a time constant > 2 second,
5.
Intermediate Range, Neutron 1 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER Flux 5
5 6.
Source Range, Neutron Flux 1 10 counts per second
< l.3 x 10 counts per second L.
7.
0vertemperature AT See Note 1 See Note 3 8.
.0verpower AT See Note 2 See Note 3 g
9.
Pressu'rizer Pressure--Low
> 1870 psig
> 1860 psig l
mR
- 10. Pressurizer Pressure--High.,
1 2385 psig.
1 2395 psig
- 11. Pressurizer Water Level--High 1 92% of instrument span 1 93% of instrument span
- 12. Loss of-Flow
> 90% of design flow per loop *
> 39% of design flow per loop *
~* Design flow is 95,000 gpm per loop.
.=.
TABLE 2.2-1 (Continued) 8 g
REACTOR TRIP SYSTE!! INSTRUMENTATION TRIP SETPOINTS
];
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E
l7% of narrow range instrument l e-
- 13. Steam Generator Water
> 18% of narrow range instrument
'E Level--Low-Low span-each steam generator span-each steam generator
.w
- 14. Steam /Feedwater Flow
< 40% of full steam flow at
< 42.5% of full steam flow at
. Mismatch and Low Steam RATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level 1,25% of narrow range instru-3.24% of narrow range instru-ment span--each steam generator ment span--each steam generator
- 15. Undervoltage-Reactor 1.2905 volts-each bus 3.2870 volts-each bus l
Coolant Pump Busses
- 16. Underfrequency-Reactor
> 56.1 Hz - each bus
> 56.0 Hz - each bus s3',
Coolant Pump Busses-17.. Turbine Trip A.
Low Trip System 1,45 psig
> 40 psig Pressure B.
Turbine Stop Valve t 1% open 1,0% open Closure 18.. Safety Injection Input Hot Applicable Not Applicable from ESF b
i9. Reactor Coolant Pump "ot Applicable Not Applicable
=
t' Breaker Position Trip.
@r+
i r
9
TABLE 2.2-1 (Continued) 5
.5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 3E NOTATION 5
[K -K "1
(T-T')+K (P-P')-f (aI)]
NOTE 1: Overtemperature AT 1 AT i 2 3
y g
1*25 Q
where:
AT,
= Indicated AT at RATED THERMAL POWER T
= Average temperature, 9F T'
= Indicated T at RATED THERMAL POWER 1 587.8 F avg P
= Pressurizer pressure, psig
'?
P'
= 2235 psig (indicated RCS norminal operating pressure)
I"lb
= The function generated by the lead-lag controller for T dynamic compensation avg 1+1 b 2
r = 25 secs, Ty &'r2
= Time constants utilized in the lead-lag controller for T,yg g 2 = 4 secs.
1 S
= Laplace transform operator (sec-1) s 8'-
a e
'.[.
.N 4
i... _....................
TABLE 2.2-1 (Continued)
Mg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5>
NOTATION (Continued) 1 E
Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops q
(no loops isolated)*
(1 loop isolated)*
1.085 K
K
(
)
K
(
)
=
=
=
3 3
3 K
0.0150 K
(
)
K I
)
=
2 2
2 K
0.000670 K
I
)
E I
)
=
3 3
3 and fi ( I) is a function of the indicated difference between top and bottom detectors
'?
of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q, - q between - 32 percent and + 9 percent, f3 (AI) = 0 l
(wher6 q Sndq are percent RATED THERMAL POWER in the top and bottom halvesofthecbrerespectivel,andqt+9 is total THERMAL POWER in b
percent of RATED THERMAL POWER k
(ii) foreachpercentthatthemagnitudeof(qkyrebu)cedby1.92percentof
-q exceeds - 32 percent, a
the AT trip setpoint shall be automatical g
its value at RATED THERMAL POWER.
5 (iii) for each percent that the magnitude of (q, - qh) exceeds + 9 percent, 2
the AT trip setpoint shall-be automaticalTy reauced by 1.77 percent of its value at RATED THERMAL POWER.
w?
E R
.
- Values dependent on NRC approval of ECCS evaluation for these operating conditions.
E
TABLE 2.2-1 (Continued) 2 E
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g
2 E-NOTATION (Continued) f
,S' q
Note 2: Overpcwer AT 1 aT [K -K T-K I~
~ 2(AIII g
4 5 1+r 35 6
l Indicated AT at RATED THERMAL POWER Where:
AT
=
g Aver % :. temperature, OF T
=
Indicated T,yg at RATED THERMAL POWER 1 587.8 F T'
=
K
'1.091
=
4 m
b 0
0.02/ F for. increasing average temperature K
=
5 0 for decreasing average temperatures K
=
5 0.00121 for T:> T'; K6 = 0 for T 1 T' K
=
6 5
33
=. The function generated by -the rate lag controller for T"V9 i
k-1+r S.
dynamic compensation-g 3
)
'3 Time constant utilized in the rate lag controller for T
=
M E
,r3,= 10 secs.,
Laplace transform. operator (sec-1) 2-P S
=
~
f (aI) = 0 for all AI 2
E Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than g
2 percent span.
G
..m___
TABLE 3.2-1
.E
- f DNB PARAMETERS
&f LIMITS E
2 Loops in Operation **
2 Loops in Operation **
q 3 Loops in
& Loop Stop
& Isolated Loop PARAMETER Operation Valves Open Stop Valves Closed 0
Reactor Coolant System T 1 592 F 3yg Pressurizer Pressure 1 2205 psig*
Reactor Coolan.t System 1
Total Flew Rate 285,000 gpm R
a
'?G 5
- Limit not applicable during either a TijERMAL POWER ramp increase in excess of 5% RATED THERMAL g
POWER per minute or a THE@iAL POWER step increase in excess of 10% RATED THERMAL POWER.
p
- Values dependent on NRC approval of ECCS evaluation for these conditions.
E i.
^>
POWER DISTRIBUTION LIMITS AXIAL POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION 1
3.2.6 The axial power distribution shall be limited by the following relationship:
[2. 20] [K(Z)]
[F (Z)]3
=
j (R))(P )(1.03)(1 + a )(1.07)
L 3
Where:
l a.
F (Z) is the normalized axial power distribution from thimble j at core elevation Z.
1 b.
P is the fraction of RATED THERMAL POWER.
c.
K(Z) is the function obtained from Figure 3.2-2 for a given core height location.
I), for thimble j, is determined from at least n=6 incore d.
flux maps covering the full configuration of permissible rod patterns above P,% of RATED THERMAL POWER in accordance-with:
n R
=
j g3 i=1 Where:
Meas p
Qi ij (Z)] Max and -[F )(Z)]gg is the maximum value of the normalized g
axial distribution at elevation Z from thimble j.in map i which had a measured reaking factor without uncertainties eas or' densification allowance of F q
j.
l NORTH ANNA - UNIT:1
'3/4 2-16
' Amendment No.. 3, l 5.176 ; 22,4 77,/19 451 L
L
_