ML20080G666
| ML20080G666 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 02/04/1984 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19277F954 | List: |
| References | |
| TAC-53214, TAC-53215, NUDOCS 8402130564 | |
| Download: ML20080G666 (22) | |
Text
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UNITED STATES 8
j NUCLEAR REGULATORY COMMISSION E
WASHINGTOls D. C. 20565 4,..... s/
DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 28 Licens? 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 (licensee) dated November 18 and supplemented Decenber 5 and 8,1983, complies with the standards and requirements of the Atonic Energy Act of 1954, as anended (the Act) and the Commis-sion's 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 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 safe'y of the public, and (ii) that such activities will be conducted in co'oliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; 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.
2.
Accordingly, the license is hereby amended by page changes to the Technical l
Specifications as indicated in the attachments to this license amendment and l
paragraph 2.C.(2) of Facility Operating License No. NPF-9 is hereby amended to read as follows:
1 8402130564 840204 PDR ADOCK 05000369 P
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(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 24 are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Tech-nical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor G. Adensam,= Chief Licensing Branch No. 4 Division of Licensing
Attachment:
Technical Specification Changes Date of Issuance: February 4, 1984 e
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4 DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION, UNIT 2 AM_ENDMENT TO FACILITY OPERATING LICENSE Anendment No. 9 License No. NPF-17 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility) Facilit3 Operating License No. NPF-17 filed by the Duke Power Conoany (licensee) dated November 18 and supplemented Decenber 5 and 8,1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as anended (the Act) and the Commis-sion's 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 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 Connission's regulaticr.s set forth in 10 CFR Chapter I; D.
The issuance nf this license amendment will not be inimical to the common defense and security or to the health and safety of the public; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Conmission's regulations and all applicable requirenents have been satisfied.
2.
Accordingly, the license is hereby anended by page chances to the Technical Soecifications as indicated in the attachments to this license anendnent and paragraph 2.C.(2) of Facility Operating License No. NPF-17 is hereby amended to read es follows:
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, I (2) Technical Specifications J
The Technical Specifications contained in Appendix A, as revised through Amendment No. 9, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Tech-nical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h
Elinor G. Adensan, Chief Licensing Branch No. 4 1
Division of Licensing
Attachment:
Technical Specification Changes Date of Issuance: February 4, 1984 i
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ATTACHMENT TO LICENSE AMENDMENT N0. 28 FACILITY OPERATING LICENSE NO. UPF-9 DOCKET No. 50-369 A"D TO LICENSE A:iEt!0 MENT WO.,9_
FACILITY OPERATING LICENSE H0. 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 a vertical line indicating the area of change. The corresponding over-leaf pages are also provided to maintain document completeness.
Amended Overleaf Pace Page III IV V
'/I 2-2 2-1 2-2a (new page) 2-5 2-6 2-9 2-10 3/4 2-9 3/4 2-9a (new page) 3/4 2-10 3/4 2-10a (new page)
B 3/4 2-4 B 3/4 2-5 i
l I
INDEX SAFETY LI!!ITS AND LIMITING SAFETY SYSTEM SETTINGS
.S. ECTION
_PAGE z.1 SAFETY LIMITS i
2.1.1 REACTOR CORE.
2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..............................
2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION...
2-2 FIGURE 2.1-lb UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN 0PERATION.............................................
2-2a FIGURE 2.1-2 (BLANK)...............................................
2'3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............
2-4 4
TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.....
2-5 Y
BASES 4
i SECTION PAGE 2.1 SAFETY LIMITS 2.11 REACTOR C0RE.................................................
B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..............................
B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS................
B 2-3 McGUIRE - UNITS 1 and 2 III
. Amendment No. 28 (Unit 1)
Amendment No. 9 (Unit 2)-
w,
INDEX LIMITING CONDITIONS FOR OPERATION A!jp SURVE1LLANCE REMREMENTS SECTION PAGE 3/4.0 APPilCAE!LIIY.
3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200 F............................
3/4 1-1 Shutdown Margin - T,yg < 200 F...........................
3/4 1-3 Moderator Temperature Coef ficient.........................
3/4 1-4 Minimum Temperature for Criticality.......................
3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown......................................
3/4 1-7 5
Flow Paths - Operating....................................
3/4 1-8 Charging Pump - Shutdown..................................
3/4 1-9 Charging Pumps - Operating................................
3/4 1-10 Borated Water Source - Shutdown...........................
3/4 1-11 Borated Water Sources - Operating.........................
3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height...,...........................................
3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D.....................
3/4 1-16 Position Indication Systems - Operating...................
3/4 1-17 Position Ir.dication System - Shutdown.....................
3/4 1-18 Rod Drop Time (Units 1 and 2).............................
3/4 1-19 Shutdown Rod Insertion Limit..............................
3/4 1-20 l
i IV McGUIRE - UNITS 1 and 2
INDEX LIMIlING CONDITIONS rR OPERATION AUD SURVEILLANCE FMQUIREMENTS S EG 10N PAGE Contrel Rod I%crtion Lir its 3/4 1-?1 FIG'JRE 3.1-1 RCD BANK INSERTION LIMITS VERidS THERMAL POWER FOUR LOOP OPERATION...
3.4 1-22 FIGURE 3.1-2 (BLANK)...............................................
3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.....................................
3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER.......
3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)......................
3/4 2-4 9
FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT.
3/4 2-5 q
3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.................................................
3/4 2-8 FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R AND R IN OPERAT10N....................
2 - FOUR LOOPS 1
3/4 2-9 FIGURE 3.2-3b UNIT 2 RCS FLOW RATE VERSUS R AND R - FOUR LOOPS 1
2 IN OPERATION.....................................
3/4 2-9a FIGURE 3.2-4 R0D BOW PENALTY AS A FUNCTION OF BURNUP..............
3/4 2-10 FIGURE 3.2-4b UNIT 2 ROD BOW PENALTY AS A FUNCTION OF BURNUP......
3/4 2-10a 3/4.2.4 QUADRANT POWER TILT RATI0.................................
3/4 2-12 3/4.2.5 DNB PARAMETERS..........................................
3/4 2-15 TABLE 3.2-1 DNB PARAMETERS..............
3/4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 RESCTOR TRIP SYSTEM INSTRUMENTATION.......................
3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION..................
3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES...
3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............
3/4 3-11 McGUIRE - UNITS 1 and 2 V
Amendment No. 28 (Unit 1)
Amendment No.
9 (Unit 2)
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVLILLANCE REQUIREMENTS SECi10N PAGE 3/4.3.2 FNGINEERED SAFETY FEATURES ACTUATION SYSTLM INSTRUMENTAlION...........
3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION....................................
3/4 3-16 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTL'ATION SYSTEM INSTRUMENTATION TRIP SETP0INTS.....................
3/4 3-25 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............
3/4 3-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........
3/4 3-34 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations.................
3/4 3-40 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS.........................................
3/4 3-41 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT i
OPERATIONS SURVEILLANCE REQUIREMENTS...............
3/4 3-43 Movable Incore Detectors..................................
3/4 3-45 Seismic Instrumentation...................................
3/4 3-46 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION...................
3/4 3-47 TABLE 4.3-4 SEISMIC HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................................
3/4 3-48 Meteorological Instrumentation............................
3/4 3-49 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............
3/4 3-50 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEI LLANCE REQUIREMENTS..........................
3/4 3-51 Remote Shutdown Instrumentation...........................
3/4 3-52 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION...........
3/4 3-53 VI McGUIRE - UNITS 1 and 2
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
[
2.1 EiFETY LIMIIS REALTOR t. ORE
- 2. e. )
The corbination of lHERMTL POWER, pressuiizer pressure, and the highest operating loop coolant temperature (T shall not exceed the limits shown in figures 2.1-1and2.1-2forfourandthhe)eloopoperation,respectively.
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 HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the require-ments of Specification 6.7.1.
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 t
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 />, and comply with the requirements of Specification 6.7.1.
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, and comply'with the requirements of Specification 6.7.1.
McGUIRE - UNITS 1 and 2 2-1
jj l. j Ol.. j..l ' l ;--l.: l::iij..
- l~
'i !!]
.] :
" FLOW PER LOOP = 97,500 gpm;.
~j:
- il
.i i:-
680
...l.;
j:!!;
. L.
ijd g
i;l g
~,:
.:d
+
- I
- I'. iil[,[I[
l UNACCEPTABLE ;
.pl Oc$
". }. :
T
- ~
OPERATION i"
g; 660 q W-
. }. -.... f.;.7%j
. pf ;' ~;.,.,
j t,.;
3.
j gI
.n gy
' t ; r.
oI pg d.:.. %
.:l g2 640
-+
O i
l-t I
-g N*3
-s~
Oy g
200g'#8'.
"I p2 D,
- i. _ _
iI
~
so 3,
y 8j g
H 620
- !.L a..... ! 860.
N I
< c:
j
, l,'
j E
g E
kl Nd!
Y oR
- i..
N,17# osi, S
^
-5
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o
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gg j
1 I.
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.. ;..:...[:
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gE
.. a ;
580
---~l-j
]$
ACCEPTABLE' l
:g g{
OPE R ATION.
!~;
- l zz WW
!j p
r i-
+-
560
-r l
t n
.I
"[:
i
.i.
- I
- J-i.:.
i.:
.p it -
-l t i-t 1-l
~l 3?l r-0 0.2 0,4 0.6 0.8 1.0 1.2 FRACTION OF R ATED THERMAL POWER l
Figure 2.1-1A REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION UNIT 1 i
l McGuire - UNITS 1 and 7.
2-2 Amendment No.28 (Unit 1)
Amendment No. 9 (Unit 2) l
665 FLOW PER LOOP = 95.500 GPM N
660 N
N P,sy,
~
645
- ^
N IJNACCEPJRBLE_
N "Qpg K
OPEltRTION 640 i
635 N N
- Pgf,
\\
c.
Ne A
\\\\
oo 620 xtg
\\
3\\
S N
N
\\\\
8' N
s\\ \\
s "5
N\\ \\\\
600 3
\\\\\\
ACCEP"RBLE 590 C PERAHCN
\\\\
585 g
580 575 g
570 0
.1
.2
.3
.4
.5
.6
.7
.8
.9 1.0 1.1 1.2 POWER (FRRCTION OF NOMINAL)
FIGURE 2.1-lb UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION McGUIRE - UNITS 1 and 2 2-2a Amendment No. 28 (Unit 1)
Amendment No.
9 (Unit 2) 4
.1
TABLE 2.2-1 5
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E
- 1. Manual Reactor Trip N.A.
N.A.
w l
- 2. Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Setpoint - 5 26% of RATED THERMAL POWER THERMAL POWER High Setpoint - < 109% of RATED High Setpoint 5 110% of RATED THERMAL POWER THERMAL POWER
- 3. Power Range, Neutron Flux,
' 5% of RATED THERMAL POWER with
< 5.5% of RATED TFER"il POWER l
High Positive Rate
's time constant 1 2 seconds with a time constant : 2 seconds
- 4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with
< 5.5% of RATED THERf'AL F0WER High Negative Rate a time constant 2 2 seconds with a time constant 1 2 seconds 7
- 5. Intermediate Range,~ Neutron 5 25% of RATED THERMAL POWER 5 30% of RATED THriRMAL POWER Flux i
5 5
- 6. Source Range, Neutron Flux
< 10 counts per second
< 1.3 x 10 counts per second
- 7. Overtemperature AT See Note 1 See Note 3
- 8. Overpower AT See Note 2 See Note 3 RS
- 9. Pressurizer Pressure--Low
~> 1945 psig
~> 1935 psig EE
.22
'10. Pressurizer Pressure--High 5 2385 psig 5 2395 psig l
55 gz
- 11. Pressurizer Water Level--High 5 92% of instrument span 5 93% of instrument span
- 12. Low Reactor Coolant Flow 1 90% of design flow per loop
- 2 89% of design ficw per loop
- EF
- Design flow is 97,500 gpm per loop for Unit 1 and 95,500 gpm per loop for Unit 2.
XX 50 o
TABLE 2.2-1 (Continued)
?
O REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 3
- 13. Steam Generator Water 1 12% of span from 0 to 30% of 2 11% of span from 0 to 30% of level--Low-Low RATED TilERMAL POWER, increasing RATED TilERMAL POWER, increasing linearly to 2 54.9% of span at to 53.9% of span at 100% of RATED 100% of RATED TilERMAL POWER.
TilERMAL POWER.
sa
- 14. Ilndervoltage-Reactor 1 5082 volts-each hus 2 5016 volts-cach bus Coolant Pumps 15..Underfrequency-Reactor 1 56.4 liz - each bus 1 55.9 Ilz - each bus Coolant Pumps 1
l
- 16. Turbine Trip a.
Low Trip System Pressure
> 45 psig 1 42 psig
}
b.
. Turbine Stop Valve Closure
> 1% open
> 1% open
- 17. Sa.9ty Injection Input N.A.
N.A.
from L5F 18.
Reactor Trip System Interlocks
-10
-11 a.
Intermediate Range Neutron Flux, P-6,
> 1 x 10 amps
> 6 x 10 amps Enable Block Source Range Reactor Trip f
b.
Low Power Reactor Trips Block, P-7 1)
P-10 Input, 10% of RATED 2 9%, 5 11% of RATED TilERMAL POWER Tl!ERMAL POWER I
2)
P-13 Input 5 10% RTP Turbire i 11% RTP Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent
TABLE 2.2-1 (Continued)
,k REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 7
7 NOTATION (Continued)
E NOTE 1:
(Continued)
Z.
v
{
P
=
Pressurizer pressure, psig, P'
2235 psig (Nominal RCS operating pressure),
=
~1 S
=
Laplace transform operator, sec and f (dI) is a function of the indicated difference between top and bottom detectors y
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 qt 9b between -36% and +9.5% for Unit 1 or between -36% and +8% for Unit 2, f (AI) = 0, where qt and q are percent RATED THERMAL POWER in the top and bottom y
b halves of the core respectively, and qt
- 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of q 9b exceeds -36%, the AT Trip Setpoint t
shall be automatically reduced for Unit 1 by 0.863% or for Unit 2 by 1.173% of its value at RATED THERMAL POWER; and EE (iii) for each percent that the magnitude of qt qb exceeds +9.5%, for Unit 1 or +8%
33@@
for UNIT 2, the AT Trip Setpoint shall be automatically reduced for Unit 1 by 0.983% or for Unit 2 by 0.901% of its value at RATED THERMAL POWER.
.E 5 CC r
3, 3, e c+
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS R
c)
NOTATION (Continued)
S A
NOTE 2:
OVERPOWER AT g
11 1 5 1
1 5
a T (y, 7 3) f AT, $ -K5(1+T S) Il+1 S) T -K6 (1 + t 5) - T"] - f (O }I 2
j Z,
4 s
i,*,,
1 5
4 4
uv
~
w As defined in Note 1, Wh~ere:
AT
=
, 'g,
j d
1 As defined in Note 1,
=
y y.
3 I
i
=,As defined in Note 1, l
As-defined in Note 1, AT
/
=
g K
5 1.0908, 4
0.02/ F for increasing average temperature and 0 for decicasing avercge rf K
=
5 temperature, g
T b S
= The function generated by the rate-lag controller for T d7"3'iC avg 1+rS5 compensation, Time constant utilized in the rate-lag controller for T,yg, 15 = 5 sec,
=
1 5 1
As defined in Note 1,
=
y.
As defined in Note 1,
=
14 0.00126/*F for T > T" and K = 0 for T 1 T",
K
=
6 6
As defined in Note 1, T
=
at RATED THERMAL POWER,
$ 588.2*F Reference T,g T"
=
As defined in Note 1, and S
=
0 for all al.
f (AI)
=
2
l I
i PENALTIES OF 0.1% FOR UNDETECTED
~ ~ - ~ ~ ~ ~ - ~ ' - ~ ~
~ ~ ~ ~
~
2 FEEDWATER VENTURl FOULING AND ACCEPTABLE n
-- a - -- -
S ME ASUREMENT UNCERTAINTIES OF l
OPE R ATION E
4G 1.7% FOR FLOW AND 4% FOR INCORE
- - - - - - - - - - - ~
m-REGION FOR 7
MEASUREMENT OF F]g ARE INCLUDED R ON 2
IN THIS FIGURE.
I l
g
/,
I 0
a-44
--I-----
- - l - - ---
t l-i
- E I
t-e o
ACCEPTABLE t- - -
'*1 1.~
'~ *I--
- E UNACCEPTABLE i 5.
"o OPERATION l
l i
i l
OPE R ATION N
C R E G I O N F O R _.. _...__ ___ __ _.. ! __ _..
8 REGION f
l l
R, & R2
-]
m i
S 42
i----
( 1.016. 42.0) 3 7, - - - -
f i
u.
1 i
i 1
- -~ ' -~~!"-
y
... _L. ____ _[_ __. __. r _. _ __ {..
.._____j._--._.
I j.
i i
O i
I i
l 1-i
--t---
7 s
. _.._.. _.... _ __ _ _. J... _... :!
i I
_ _ _j _ q___
.__L e
m 49 o
I (1.0. 39.702)j C
i
- 1...._ _ _
s I
- ' ACCEPTABLE OPERATION
{
r_____.___
i i.
-REGION FOR590% RTP t
i 7.
38
- - -----i
- t- -
t.
i I' -
3 yy (1.0,37.717) !
au
]
-h t
s I
I-1 -- -
-+
-l..--
ma gg
.. UNACCEPTABLE
_ h _..___
p go OPERATION e&
REGION l
[ ~~ l - ~ ~ ~ --
--j.
gg gg 0.90 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 w
- j R
= Fyg/1.49 l 1.0 + 0.2 (1.0 - P)l 3
((
R = R /l1 - RBP(Bu)l 2
j ee 3
FIGURE 3.2-3A UNIT 1 RCS TOTAL FLOWRATE VERSUS R AND R - FOUR LOOPS IN OPERATION i
2
u.
11..,, m.~+~.
.m, 7..
an
.e7,.
g.fn itz; : n;. :g3:_g,ef++ ++*~.
. y
++nn 7
+++-
n..
++++ :m i.:n..m.ua re r"n.;
-,~.
y
.,1
. :n; g.
m.
2 n
.... eu
..; g m
e, e
m,t::: tn t:ttutt:
o i
7 g ng
- .tw oit "++
fiHn.t: :n:
C i
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w i rr 0 F& 0 5 10 15 20 25 30 3 REGION RVERAGE BURNUP (10 MWD /MTU) CC 3 3 eo FIGURE 3.2-4b UNIT 2 ROD 80W PENALTY AS A FUNCTION OF BURNUP ~~ vv 1 l
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALFY RISE HOT CHANNEL FACTOR (Continued) The control rod insertion limits of Specifications 3.1.3.5 and c. 3.1.3.6 are maintained; and d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F will be maintained within its limits provided Conditions a. through H
- d. above are maintained.
As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and F may be " traded off" against one another (i.e., a low measured RCS flow H N rate is acceptable if the measured F is also low) to ensure that the calcu-3H lated DNBR will not be below the design DNBR value. The relaxation of F as AH a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts y for F less than or equal to 1.49. This value is used in the various accident g analyses where F influences parameters other than DNBR, e.g., peak clad tem-g perature, and thus is the maximum "as measured" value allowed. R, as defined, 2 allows for the inclusion of a penalty for Rod Bow on DNBR only. Thus, knowing the "as measured" values of F and RCS flow allows for " tradeoffs" in excess H of R equal to 1.0 for the purpose of offsetting the Rod Bow DNBR penalty. Fuel rod bowing reduces the value of DNB ratio. Credit is available to partially offset this reduction. This credit comes from a generic or plant specific design margin. For McGuire Units 1 and 2, the margin used to partially offset rod bow penalties is 9.1 percent. This margin breaks down as follows: 1) Design limit DNBR
- 1. 6%
2) Grid spacing K 2.9% s 3) Thermal Diffusion Coefficient 1.2% 4) DNBR Multiplier 1.7% i 5) Pitch Reduction 1.7% For McGuire Unit 2, the margin used to partially offset rod bow penalties is 5.9 percent with the remaining 3.2 percent used to trade off against measured flow being as much as 2 percent lower than thermal design flow plus uncertain-ties. The penalties applied to F to account for rod bow (Figures 3.2-4 Unit 1 H i and Unit 2) as a function of burnup are consistent with those described in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5, 1979 with the difference being due to the amount of margin each unit uses to partially offset rod bow penalties. McGUIRE - UNITS 1 and 2 B 3/4 2-4 Amendment No. 28 Unit 1 Amendmont No. 9 Unit 2
i e POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHAN!EL FACTOR w RCS FLCW PATE AND NUCLEfR ENTHALPY RISE H0f CHANNil FACTrs (Continued) When an F measurement is taken, an ailowance for both experimental error q and manufacturing tolerance must,be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. When RCS flow rate and F are measured, no additional allowances are g necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. j Measurementerrorsof1.7%forRCStotalflowrateand4%forFhhavebeen allowed for in determination of the design DNBR value. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate 'r indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-3. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during 4 power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. A limit of 1.02 was selected to provide an allowance far the q uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc- ] tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q the maximum allowed power oy 3% for each percent of tilt in excess of 1.0. McGUIRE - UNITS I and 2 B 3/4 2-5 Amer.dment No. 28(Unit 1) Amendment No. 9(Unit 2) .}}