ML20154P764
| ML20154P764 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/19/1988 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20154P768 | List: |
| References | |
| NUDOCS 8806060071 | |
| Download: ML20154P764 (14) | |
Text
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UNITED STATES t
NUCLEAR REGULATORY COMMISSION g
WASHING TON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. NPF-9 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-9 filed by the Duke Power Company (the licensee) dated October 29, 1985, as supplemented August 25, 1986, May 26, 1987, and January 19, 1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, 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 I
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.
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8806060071 080519 PDR ADOCK 05000369 P
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Accordingly, the license is hereby amended,by page changes to the. Technical 2.
Specifications as indicated in the attachments to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-9 is hereby~,
amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 84, are hereby incorporated into the license.
The licensee shall operate the facility in accordance with the Technical Specifications.and the Environmental Protection Plan.-
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION.
Original signed by:
David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II
Attachment:
Technical Specification Changes Date of Issuance: }iay 19, 1988 0FFICIAL RECORD COPY
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p" UNITED STATES 3"
NUCLEAR REGULATORY COMMISSION k
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DUKE POWER COMPANY DOCKET N0. 50-370 McGUIRE NUCLEAR STATION, UNIT 2 A_MENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. NPF-17 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, L' nit 2 (the facility) Facility Operating License No. NPF-17 filed by the Duke Power Company (the licensee) dated October 29, 1985, as supplemented August 25, 1986, May 26, 1987, and January 19, 198S, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, 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 Commission's regulations set forth in 10 CFR Chapter I; 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 anlendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly,.the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-17 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 65, are hereby incorporated into the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:
David B. Matthews, Director Project Directorate II-3 Division of' Reactor. Projects-I/II
Attachment:
Technical Specification-Changes Date of Issuance: May 19, 1988 l
1 i
l 0FFICIAL RECORD COPY
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ATTACHMENT TO LICENSE AMENDMENT NO. 84 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET N0. 50-369 AND TO LICENSE AMENDMENT N0.165 FACILITY OPERATING LICENSE N0. NPF-17 DOCKET N0. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
The corresponding over-leaf page is also provided to maintain document completeness.
Amended Page Overleaf Page 2-5 2-8 2-9 2-11 B 2-4a (new page)
B 2-5 3/4 3-1 3/4 3-2 3/4 3-9
TABLE 2.2-1
.c S
l 5
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l
A FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES c-l 5
- 1. Manual Reactor Trip N.A.
N.A.
- 2. Power Range, Neutron Flux Low Setpoint 5 25% of RATED Low Setpoint 1 26% of RATED g
THERMAL POWER THERMAL POWER 5
High Setpoint 5 109% of RATED High Setpoint 5 110% of RATED m
THERMAt POWER THERMAL POWER
- 3. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 5 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 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Negative Rate a time constant 2 2 seconds with a time constant 1 2 seconds 7
- 5. Intermediate Range, Neutron 1 25% of RATED THERMAt. POWER
$ 30% of RATED THERMAL POWER u'
Flux
- 6. Source Range, Neutron Flux 5 105 counts per second 5 1.3 x 105 counts per second
- 7. Overtemperature AT See Note 1 See Note 3**
l yy
- 8. Overpower AT See Note 2 See Note 4**
l pg
- 9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig
- 10. Pressurizer Pressure--High 1 2385 psig 5 2395 psig zz
- 11. Pressurizer Water Level--High -< 92% of instrument span 5 93% of instrument span
- 12. Low Reactor Coolant Flow 1 90% of design flow per loop
- 188.8% of design flow per loop
- EE g
- Design flow is 97,220 gpm per loop.
- Prior to removal of each unit's RID bypass manifold, note 3a is applicable.
l
TABLE 2.2-1 (Continued)
E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
NOTATION E
] NOTE 1: OVERTEMPERATURE AT 1+t S 1
AT (1
) (y 1 3) $ AT {K -K2 (1 +
$)[T(y.
3)-T'] + K (P-P') - f (AI)}
1 7
T3 g
y 3
y m
Where:
AT Measured AT by RTD Manifold Instrumentation,
=
I b
Lead-lag compensator on measured AT,
=
g 13, r2
= Time constants utilized in the lead-lag controller for AT, 11 1 8 sec., T2 5 3 sec.,
1 E
Lag compensator on measured AT,
=
y.
r3 Time constants utilized in the lag compensator for AT, 13 $ 2 sec.*
l
=
AT, Indicated AT at RATED THERMAL POWER,
=
~ 1.200, K
K
= 0.0222 2
&R 1+1S4 hh 1+rS The function generated by the lead-lag controller for T dynamic compensation,
=
s avg Time constants utilized in the lead-lag controller for T**9,
=
o,cn 14 13
{
t 1 28 sec, 13 5 4 sec.,
CC 11 T
=
Average temperature, F,
or 1
SC Lag compensator on measured T,yg,
=
y.
3 i
2 TABLE 2.2-1 (Continued)
S S
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS A
NOTATION (Continued) c 3
NOTE 1:
(Continued)
--e Ts Time constant utilized in the measured T 1 g compensator, is 5 2 sec*
=
avg o,g-T' 5 588.2 F Reference T at RATED THERMAL POWER,
=
avg K
0.001095, 3
P
=
Pressurizer pressure, psig, P'
2235 psig (Nominal RCS operating pressure),
=
Laplace transform operator, sec 1,
S
=
and f (AI) is a function of the indicated difference between top and bottom detectors e
i 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 9 between -29% and +9.0%; f (AI) = 0, where q t
b y
t and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q
- 9 t
b
(( -
is total THERMAL POWER in percent of' RATED THERMAL POWER; 22 l
k(
(ii) for each percent that the magnitude of q
~9 exceeds -29%, the AT Trip Setpoint t
b 55 shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and zz
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(iii) for each percent that the magnitude of q q exceeds +9.0%, the AT Trip Setpoint t
b m oo shall be automatically reduced by 1.50% of its value at RATED THERMAL POWER.
22 1
c TABLE 2.2-1 (Continued)
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gg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
- o NOTATION (Continued)
E
((
T
=
As defined in Note 1,
((
T"
$ 588.2 F Reference T at RATED THERMAL POWER,
=
avg a
f, S
As defined in Note 1, and
=
f (AI) 0 for all AI.
=
2 Note 3:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.6% of Rated Thermal Power.
Note 3a:
The channel's maximum Trip Setpoint shall not exceed its computed. Trip Setpoint by more than 2%.
U Note 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.2%
of Rated Thermal Power.
NN i
aa-ft&
a" en 28 EL EL ee i
s LIMITING SAFETY SYSTEM SETTINGS BASES (With RTD Bypass System Installed)
Overtemperature AT The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the Pressurizer High and Low Pressure trips.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion.
With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Overpower AT The Overpower Delta T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for overtemperature delta T protection, and-provides a backup to the High Neutron Flux trip.
The Setpoint j
is automatically varied with:
(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution, to ensure that 1
the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT trip provides protection to mitigate the consequences of various size steam i
breaks as reported in WCAP 9226, "Reactor Core Response to Excessive Secondary i
Steam Break."
i i
McGUIRE - UNITS 1 and 2 B 2-4a Amendment No. 84(Unit 1)
Amendment No. 65(Unit 2)
d a
LIMITING SAFETY SYSTEM SETTINGS BASES (With Bypass System Removed; RTDs in Thermowells)
Overtemperature AT The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to thermal delays associated with the RTDs mounted in thermowells (about 5 seconds),
and pressure is within the range between the Pressurizer High and Low Pressure trips.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature oetectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Overpower AT l
The Overpower Delta T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for overtemperature delta T protection, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for instrumentation delays associated with the loop temperature detectors, and (31 axial power distribution, to ensure that the allowable heat generation rate (kW/tt) is not exceeded.
The Overpower ATtrip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP 9226, "Reactor Core Response to Excessive Secondary Steam Break."
McGUIRE - UNITS 1 and 2 B 2-5 Amendment No. 84(Unit 1)
Amendment No. 65(Unit 2)
s 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System Instrumentation channels and interlocks of Table 3.3-1 shall'be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICABILITY:
As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System Instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one train such that both trains are tested at least once per 36 months and one channs, per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.
4.3.1.3 The response time of RTDs associated with the Reactor Trip System shall be demonstrated to be within their limits (see note 2 to Table 3.3-2) at least once per 18 months.
McGUIRE - UNITS 1 and 2 3/4 3-1
' Amendment No84(Unit 1)
Amendment No65(Unit 2)
c 1
TABLE 3.3-1 2
S C
REACTOR TRIP SYSTEM INSTRUMENTATION
~
A MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE c
2 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.
1 2
1, 2 1
y 2
1 2
3*, 4 *, 5*
10 2.
Power Range, Neutron Flux - High 4
2 3
1, 2 2
Setpoint
. Low 4
2 3
1,,,,
2 2,
Setpoint 3.
Power Range, Neutron Flux 4
2 3
1, 2 2
High Positive Rate g"
4.
Power Range, Neutron Flux, 4
2 3
1, 2 2
High Negative Rate m
e 5.
Intermediate Range, Neutron Flux 2
1 2
l
,2 3
6.
Source Range, Neutron Flux a.
Startup 2
1 2
2,,
4 j
-b.
Shutdown 2
1 2
3*,
4*, 5*
10 c.
Shutdown 2
0 1
3, 4, and 5 5
3
,,___s
'm u
?-
7.
Overtemperature AT x
Four Loop Operation 4
2 3
1, 2 6
Three Loop' Operation
(**)
(**)
(**)
(**)
(**)
s i
w N
i a
L TABLE 3.3-2 x
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES A
FUNCTIONAL UNIT RESPONSE TIME
]
1.
Manual Reactor Trip N.A.
[
2.
Power Range, Neutron Flux
$0.5 second (1) l s"
3.
Power Range, Neutron Flux, High Positive Rate N.A.
4.
Power Range, Neutron Flux, High Negative Rate 10.5 second (1) l S.
Intermediate Range, Neutron Flux N.A.
6.
Source Range, Neutron Flux N.A.
t 7.
Overtemperature AT
$10.0 seconds (1)(2)(3) 8.
Overpower AT 510.0 seconds (1)(2)(3) 9.
Pressurizer Pressure--Low
<2.0 seconds 10.
Pressurizer Pressure--High
$2.0 seconds
_ gg ee y E 'll.
Pressurizer h ter Level--High N.A.
i E$
ee EE (1) Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
&E (2) The 5 10.0 second response time includes a 5.5 second delay for the RTDs mounted in thermowells.
(3) The < 10.0 second response time is applicable to each unit only after the RTD bypass manifold is mm E E, removed; until then the value 1 8.0 sec.
ee i
.