ML20108B809
| ML20108B809 | |
| Person / Time | |
|---|---|
| Site: | North Anna  (NPF-07-A-041, NPF-7-A-41) |
| Issue date: | 10/15/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20108B815 | List: |
| References | |
| NUDOCS 8411160188 | |
| Download: ML20108B809 (13) | |
Text
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UNITED STATES
+
E $.h v,(f j NUCLEAR REGULATORY COMMISSION
- p 7#(: j WASHINGTON, D. C. 20555 '
hff VIRGINIA ELECTRIC AND POWER COMPANY-OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE-
- Amendment No. 41 License No. NPF-7 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, and July 11, 1983) and September 29, 1983, comply with as amended (the Act)quirements of the Atomic Energy Act of 1954, the standards and re 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 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 activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements i
I have been satisfied.
8411160188 841015 PDR ADOCK 05000339 P
r
. l 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.C.(2)iof 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. 41, are hereby. incorporated in the. license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective upon the date of its issuance.
FOR THE NOCLEAR REGULATORY COMMISSION
/ -
-s
/
./ James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: 0ctober 15, 1984 I
r: -
,'l ;
l ATTACHMENT TO LICENSE AMENDMENT
~
AMENDMENT NO. 41 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 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.
Pages 2-2 2-6 i
2-8 2-9 10 3/4 2-16 4
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4 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 (TFigures 2.1-1 for 3 -loop operation an8V shall not exceed the. limits shown in APPLICABILITY: MODES 1. and 2.
-ACT'"N:
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 />.
9 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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' NORTH ANNA - UNIT 2 2-1
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l NORTH ANNA - UNIT 2 2-2 Amendment No. 29, 41 l
l 4
SAFETY LIMITS AND.IMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS
- REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.11 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 statement 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.
6 NORTH ANNA - UNIT 2 2-5
I TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0iNTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES-1.
Manual Reactor Trip Not Applicable Not applicable 2.
Power Range, Neutron Flux Low Setpoint - 1 25% of RATED Low Setpoint - 1 26% of RATED THERMAL POWER THERMAL POWER.
High Setpoint - 1 109% 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 1 2 seconds With a time constant 1 2 seconds.
4.
Power Range, Neutron Flux,
< 5% of RATED THERMAL POWER with.
< 5.5% of RATED THERMAL. POWER High Negative Rate a time constant > 2 seconds with a time constant 1 2 seconds.
5.
Intermediate Range, Heutron 1 25% of RATED THERMAL POWER 5 30% of RATED THERMAL POWER Flux 6.
Source Range, Neutron Flux
< 10 counts per second
< l.3 x 105 counts per second 5
7.
Overtemperature AT See Note 1.
See Note 3 8.
Overpower AT See Note 2
~See Note 3-9.
Pressurizer Pressure--Low
> 1870 psig
> 1860 psig 10.
Pressurizer Pressure--High 1 2385 psig i 2395 psig 11.
Pressurizer Water Level--High s 92% of instrument span 1 93% of instrument span 12.
Loss of Flow
> 90% of design flow per. loop *
> 89% of des'ign flow per. loop
- ODesign flow is 95,000 gpm per loop.
s.
r TABLE 2.2-1 (Continued).
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Ed REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ji FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 13. Steam Generator Water 1 18% of narrow range instrument 1 17% of narrow range' instrument SE Level--Low-Low span-each steam generator span-each steam generator G
- 14. Steam /Feedwater Flow
$ 40% of full steam flow at.
5 42.5% of full steam flow at n3 Mismatch and Low Staam RATED THERMAL POWER-coincident.
RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water levei_
1 25% of narrow range instru-1,24% of narrow range'instru-ment span--each steam generator ment span--each steam generator
- 15. Undervoltage-Reactor 1 2905 volts-each bus.
1 2870 volts-each bus Coolant Pump Busses
- 16. Underfrequency-Reactor 1 56.1 Hz - each bus 1 56.0 Hz - each bus Coolant Pump Busses n,
- 17. Turbine Trip A.
Low Trip System 1 45 psig 1 40 psig Fressure B.
Turbine Stop Valve 1 1% open 1.0% open closure
- 18. Safety Injection Input not Applicable
'Not' Applicable from ESF
- 19. Re' actor Coolant Pump Not Applicable Not Applicable Breaker Position Trip
TABLE 2.2-1 (Continued)
~
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
'l NOTATION NOTE 1:
Overtemperature AT 1 AT, [K -K2 3
j 1
(T-T ')+K (P-P ')-f (AI)]
j 1+t S 2,
where:
AT, Indicated AT at RATED THERMAL POWER
=
Average temperature, "F T
=
T' Indicated T,yg at RATED THERMAL POWER 1 587.8 F l
=
Pressurizer pressure, psig P
=
f 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
Time constants utilized in the lead-lag controller for T-t = 25 secs, t
t =
1 2
- 8 I = 4 secs.
2 Laplace transform operator (sec-1)
~
S
=
t e
I I.
TABLE 2.2-1 (Continued)'
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REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS HOTATION (Continued)
Operation with 3 loops Operation with 2 Loops Operation with 2 Loops-g (no loops isolated)*-
(1 loop isolated)*,
(
)
LK
=
(-
)~
l
]
1.085 K
K
=
=
j j
j C
(
)
K 0.0150 K
I'
)
I.
K
=
q 2
2 2
N 0.000670 K
+ (
)
K3'"'
(
)
K
=
l 3
3 and fj (AI) 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:
(1) for qt -9b between - 32 percent and + 9 percent, f) (AI) =.0 l
{
(where qt and qb are p'ercent RATED THERMAL POWER in'the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in i
percent of RATED THERMAL POWER).
[
(ii) for each percent that the magnitude of (qt - 9 ) exceeds 32 percent, l'
b E
the AT trip setpoint shall be automatically reduced.by 1.92 percent'of l
l its value at RATED THERMAL POWER.
n (iii) for each percent that the magnitude of (qt 9 ) exceeds + 9 percent, I
z b
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the AT trip setpoint shall be automatically reduced by.1.77 percent of l
5 its value at RATED THERMAL POWER.
g C
- Values dependent on NRC approval of ECCS evaluation for these operating conditions.
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TABLE 2.2-1 (Continued) j REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued) 2 O
g T3g,T-K6 (T-T")-f (AI)3 Note 2:
Overpower AT 1 AT ]K -K5 2
g 4 g
9 EE e
e Indicated AT at RATED THERMAL POWER where:
AT
=
E E
j Average temperature, F G
T
=
Indicated T at RATED THERMAL POWER 1 587.8 F.
N T"
=
3yg f
1.091 l
K
=
4 0.02/ F for increasing average temperature' i
i K
=
5 m
0 for decreasing average temperatures L
K
=
5 o
0.00121 for T > T";
K = 0 for T T"
K
=
6 6
S T3 The function generated by the rate lag controller for Tavg
=
1+T S 3
dynamic compensation Time constant u'tilized in the rate lag controller for T i
3"
'3
=
3 = 10 secs.
avg T
Laplace transform operator (sec-j)
S
=
f (aI) = 0 for all AI
=
2
,o M
Note 3:
The channel's maximum trip point shall not exceed its computed trip point by more -than,
[
2 percent span.
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POWER DISTRIBUTION LIMITS I
DNB PARAMETERS
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LIMITING CONDITION FOR OPERATION 3.2.5. The following DNB related parameters shall be maintained within the -
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% 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:
s With any of the above pac. deters 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 5% 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 />.
4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.
NORTH ANNA - UNIT 2 3/4 2-15 i
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TABLE 3.2-1
{
g c
- o k
DNB PARAMETERS E
B LIMITS 5
2 Loops in Operation **
2 Loops in Operation **
a g
s 3 Loups in
& Loop Stop
& I".olated Loop n
Operation _
Valves Open Stop Valves Closed PARAMETER g
l B
Reactor Coolant Sysi.si T 5592 F
{
avg
>2205 psig*
E Pressurizer Pressure l
p
>285,000 gpm 5
Reactor Coolant System Total Flow Rate P
I E
R 1
a Y
L I
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERM per minute or a THERMAL POWER step increase in excess of 10% RATED TiiERMAL POWER.
h E
- Values dependent on NRC approval of ECCS evaluation for these conditions.
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