ML20082P587
| ML20082P587 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  (DPR-062) |
| Issue date: | 11/28/1983 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082P592 | List: |
| References | |
| DPR-62-A-083 NUDOCS 8312090089 | |
| Download: ML20082P587 (25) | |
Text
,
[ga atog'o, UNITED STATES
!y3w
'j NUCLEAR REGULATORY COMMISSION
- y WASHINGTON, D. C. 20555
/
- ..+
i I
CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMEN'DMENT TO FACILITY OPERATING LICENSE
's i
Amendment No. 83 License No. DPR-62 1.
The Nuclear Regulatory Comr ission (the Comission) has found that:
A.
The application for amendment by Carolina Power & Light Company (thelicensee)datedJuly 29, 1983, complies 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; B.
The facility will operate in conforrgi_ty,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 regulat ons; 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 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.
s 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this' licend amendment,
~
and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:
i 831N90089 831128 PDR 3 DOCK 05000324 p
_. 2 2.
Technical Spee.ifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 83, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
t.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'\\,
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications
-Date of Issuance: November 28, 1983
- 4mW 4
i
l l
ATTACHMENT TO LICEriSE AMDMPIT *:3. 83
. FACILITY OPERATING LICENSE NO. OPR-62.
DOCKET NO. 50-324
~
Revise the Appendix A Technical Specifications as indicated below. The changed areas are indicated by vertical lines.
Rem'ove' Insert IV IV 3/4 1-6 3/4 1-6 3/4 1-7 3/4 1-7 3/4 2-1 thru 3/4 2-15 3/4 2-1 thru 3/4 2-15 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2.-3 B 3/4 2-3 8 3/4 2-5
-B-3/4 2-5 6
(
e uun
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY..............................................
3/4 0-1 3/4.1 REACTIVITY' CONTROL SYSTEMS 3/4.1.1 SH UIDOWN MARGI N..........................................
3/4 1-1 S/4.1.2 REACTIVITY AN0MALIES.....................................
3/4 1-2 3/4.1.3 4'ONTROL RODS N.
Cont'rol Rod Operability..................................
3/4 1-3 Ccntrol Roi Maximum Scram Insertion Times................
3/4 1-5 Control Rod Average Scram Insertion Times................
3/4 1-6 Four Control Rod Group Insertion Times...................
3/41-7 Control Rod Scram Accumulators...........................
3/4 1-8 Control Rod Drive Coupling...............................
3/41-9 s
Control Rod Position Indication..........................
3/4 1-11 Control Rod Drive Housing Support........................
3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod W rth Minimizer......................................
3/4 1-14 o
Rod Sequence Cont rol System..............................
3/4 1-15 Ro d Bl o ck No ni t o r........................................
3/4 1-17 3 /4.1.5 STANDBY LIQUID CONTROL SYSTEM............................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATIONN RATE..............
3/42-1 3/4.2.2 APRM SETP0INTS.........................:..........;.....
3 3/42-8 3/4.2.3 MINIMIN CRITICAL POWER RATI0.............................
3/4 2-9 3/4.2.4 LINEAR HEAT GENERATION RATE..............................
3/4 2-15 3RUNSWICK - INIT 2 IV Amendment No. 83
~__
REACTIVITY CONTROL SYSTEMS CONTROL ROD AVERAGE SCRAM INSERTION TIMES LIMITING CONDITIONS FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods from the fully,withdr. awn.pcsition, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:
Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 46 0.31
-i 36 1.05 N
26 1.82 y
6 3.37 APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With the' average scram insertion time exceeding any of the above limits, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />..
SURVEILLANCE REQUIREMENTS s 4.1.3.3 All control rods shall.be demonstrated OPERABLE by scram time testing fros the fully withdrawn position as required-by-Surveillance Requirement j
4.1.3.2.
2 t
BRUNSWICK - UNIT 2 3/4-1-6 Amendment.No. 83
p
[
~
REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES P
i.
LIMITING CONDITION FOR OPCRATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a twc-by-two array, based on deenergization of the scram pilot valve solenoids as tim'e zero, shall not exceed any of the following:
Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 46 0.33 36 1.12
.c'N.
26 1.93 6
3.58 APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
l ACTION:
With the average scram insertion times of control rods exceeding the above limits, operation may c,ontinue and the provisions of Specification 3.0.4 are not-applicable provided:
a..
The control rods with the slower than average scram insertion times s
are declared inoperable, b.
The requirements of Specification 3.tr3d are satisfied, and c.
The Surveillance Requirements of Specification 4.1.3.2.c are performed at least once per 92 days when operation is continued with three or more control rods with slow scram insertion times.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.4 All control. rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement l
4.1.3.2.
2 BRUNSWICK - UNIT 2 3/4 1-7 Amendment No.
83
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT CENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All' AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR's) for each type of fuel as a~ function of AVERAGE PLANAR EXPOSURE shall not exceed the
- following 'livul'ts:'
During two recirculation loop operation, the limits are shown in a.
I Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6.
l APPLICABILITY:.OPERATIOf'AL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THER!fAL POWER.
,.s WitY an APLHCR exceeding the limits of Figures 3.2.1-1, 3.2.1-2, ACTION:
3.2.1-3, 3.2.1-4, 3.~2.1-5, and 3.2.1-6, initiate corrective action within 15 li minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% 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 REOUIREMENTS s
4.2.1 All APLHGR's.shall be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, I~.
1-5, and 3.2.1-6:
~~
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL P0a'ER' increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.
operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
t BRIJNSWICK - INIT 2 3/42-1 Amendment No. 83
ll 6
9 0
0 0
0
\\
4 w
0 0
0 5
3
~
.s 9
0 0
1 0
0 3
0 3
N.
/0 I
0
{
0 T
0
)
A 5
t E
E 2
/
H R
IM
)t RRS M
)AGO
(
8E P
N XNlJ 1
X E
8I PE R
(LA 1
2
,x 0
U MR 1
0 S
lR(A 1
0 O
iA N
0 P
f7NEA 2
2 XE 2ATL DLAP 3
8PR R
E E,
A EENG R
4 PGOA U
M YAI R G
2 L
I 2
TRTE 0
P EAV F
1 0
LVRA 0
E EAE I
- l L
5 G
U NS BF4 A
F4EU 1
00 R
fl GS I
S 1
E M
R 1 T V
S t
I E
I f A A
X V
MiR A
RGE M
e EEP PRO 0
0 0
1s 0
1 2
1 0
r 0
0 5
9 1
1
/
0 1
0 9
'1 1
s 6O $r gaa_
v gg>< fE {
EEy.S l
w$ ?"
g$g 5 5
I
\\
11l lljl 11 6
0 9
0 0
0 4
2 N 0
1Q 0
0
, 0 5
3 9
0 0
1 0
0 3
0 3
/
/
5%
)
T t
A 1
0 /
E E
1 0 d H
R d
)U 0
5 M
RRS 2 (
)AGO 8El o E
XNuX R
8I PE U
(LA i
\\
O A
0 1
P A
N t
f 0 X 7NE 1 2
0 E 2AT1 2
1, 0
DL AP 3
0 R 8PR 2 A E
e N
EENG R
A PGOA U
YAI R LP TRTE G
2.N i
EAV e
E LVRA G
EAE 2
\\
0 A U
NS 1
0 R F4EU 0 E UGS
^
tL 5 V M
R A
I E
BFN 1
OO X
V I
A SSNU M
I
( I MI R RGE EEP PRO 0
s H
0 0
M 0
1 I
' [
0 0
0 8
5 1
2 1
1 1
~
0 1
.I 9
1 I 1
=
sv N5 8oohg yh b!=;
g8 i
3O t
3Gg gi
~
uN m&
pQ@4 3
[
1 l
ll1 ll' i,
,l Z
q I
0 g
0 0
0 4
\\
1 0
0 1
L 0
0 5
3 7
0N
'0 1
0 0
0 3
f 3
/
1 0
0
)
0 t
T 5 /
)A 2
d E
E i
R M
8 H R
X
)U
(
RRS 8
AGO
(
E E
l P
R 1
NUX 1
U 5I PE 3
0 S
6LA 0
O 2
MR 1
0 P
BR(A 0
X RA N
2 2
E DNEA 3
8ATL R
AP L
1 A
EPR E
P E
R I
R YENG U
0 TGOA G
L 2
P AI R I
1 0
LRTE F
0 E
EEAV 0
\\
G UVRA L
5 A
FAE 1
OO R
NS I
F 4EU S
I SWT V
UGS i
A M
R I
\\
Ml R I
E RGE X
V EEP A
PRO 0
M 0
0 0
0 2
1 4
1 9
1
_1 0
0
/
0 5
0 2
1 0
9 1
1~
1 md%
mB 6C g0 qE 8z3 g3' iyEk 5sg d
u 3GebQ w # 7#
lgy 5.
O lll
'I l lIlll l!
llIIi1j(
Il l
lJ i
8 9
0 0
0 s
0 N
4 N
1 4.N 0
0 1
0 0
5
\\
3 N
1 h
0 0
0 0
3 j
/ g' 0
)
0 t
0 /
5 l
T M
A 2
)E E
O R H R
s 8
)U E
X RRS 8
R (AGO 8
U El P i
S 3NUX 1
0 O 8I PE 1
0 P 2L 1
R R (W R 0 X B
A 2
0 E 2
DA W
R 8NE 3
A ATU N
ELAP E
A PPR R
L Y
E U
P TENG G
GOA I
0 E LAI R 0
G ERT"'
F
)
!i 0 A UEAV E
9, L
5 R
FVRA BFN 1
E AE 1
OO V
NS I
1 S
'I A
4 EU l
S4 UGS UA M
R I
MI R I
E 0
RGE X
V 2
EEP A
0 1
PRO M
0 0
0 1
0 0
0 5
1 1
)'?
1 1
0 p
9 i
9j* i,$ha5* "hr hz]
i ln
- g 55E g E e8X&::<-
u c sE 3LpQ@4 oc" r
l.
llli!
llll llll ll
5 E
%)
15 g
2
'7 g
/
~
-4 12.1 12.1 N
12-N f
1,g d
'9 11.6 s
v 11.5 11.3 I
7 15
'\\
5 as to t
'f h
10
~b
\\
PERMI iSIBLE REGIO 1 OF fI
\\
OPFRATION g
\\
- 9.4 s
8 9-
~
p 0
5000 10000 15000 20000 25000 30000 35000 40000 PLANAR AVERAGE EXPOSURE (mwd /t) z FUEL T(PE P8DRB265H (P8X8R)
MAXIMtJ4 AVER /CE PLANAR LINEAR llEAT m
GENERATION RATE (MAPUlGR)
VERSUS Pl.ANAR AVERAGE EXPOSURE FIGURE 3.2.1-5
0 0
9 0
0 0
+
f
~
~
~
~
7 0
9 0
L 0
5 3
3 0
0 1
0 0
0 3
\\
T
/
A 0
E j
[
)
l E
1 0
t
)
R l
1 0
/
RR)U 0
d 8ARS 5
W XEGO 2
M 8N P
(
PI UX (LPE 6
4 E
A R
l RMR 1
l
\\
U A(A i
t 7
S 8N N
O 2AEA 2
1N 1
0 P
BLTL 3
0 X
RPAP 0
E D
R E
h 0
8E E
R 2
E PGNG U
G AOA G
A ERI R I
R PETE F
E YVAV V
TARA A
E LNIDU S
R El 0
A UMGS 0
N FI R
0 A
X E
- 0. -,,
E 5
L A
V L
1 P
M 2
BFNOO 1
I S
I i
l '
- OA I
MI R RGE EEP 0
PRO 0
0 0
0 1
2 1
' /
0 0
0 4
5 7./
1
/
1 0
1 0
9 1
1 i
U
., ~
bO 2~.
(
0 2
t Qh4 5-k22g Eqn
, ug m g
ll
/
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow-biased APRM scram trip setpoint (S) and rod block trip setpoint (SRB) shall be established according to the following relationships:
'S 1 (0.66W + 54%) T SRB i (0.66W + 42%) T where:
S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow in percent of rated flow,
.T = Lowest value of the ratio of design TPF divided by the MTPF
.'bbttained for any class of fuel in the core (T 1 1.0), and Design TPF for: P8 X 8R fuel = 2.39 8 X 8R fuel = 2.39 8X8 fuel = 2.43 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
-ACTION:
With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and SRB are within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or reduce THERMAL POWER to less than 25% 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 />.
SUDIEILLANCE REOUIREMENTS 4.2.2 Ihe }ffPF for each class of fuel shall be determined, the value of T calculated, and the flow-biased APRM trip setpoint adjusted, as required:
At leasb once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a.
b.
- Within12hoursaftercompletionofaTHERMALPOWERincreaseohat least 15% of RATED THERMAL POWER, and c.
. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MTPF.
t BRUNSWICK
'JNIT 2 3/42-8 Amendment No. 83'
l POWER DISTRIBUTIOM LIMITS 3 /4. 2._3 MINIMUM CRITICAL POWER RATIO LIMITING CON 5ITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shown in Figure shall be.cqual,co or greater than the MCPR limit times the Kf 3.2.3-1, provided that the end-of-cycle recirculation pump trip system is OPERABLE per specification 3.3.6.2, with:
If ODYN OPTION A analyses are in effect, the MCPR limits are a.
listed below:
l.
MCPR for 8x8 fuel
= 1.29 2.
MCPR for 8x8R fuel = 1.27 3.
MCPR for P8x8R fuel = 1.29 b.
If ODYN OPTION B analyses are in ef f'ect (refer to Specification 3.2.3.2), the MCPR limits are listed below:
1.
MCPR for 8x8 fuel
= 1.29 2.
MCPR for 8x8R fuel = 1.21 3.
MCPR for P8x8R fuel = 1.22 APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER ACTION:
With the end-of-cycle recirculation trip systen inoperable per a.
Specification 3.3.6.2, operation may continue and the provisions of Specification 3.0.4 are not applicable with the follo ring MCPR limit adjustments:
1.
Beginning-of-cycle (BOC) to end-of-cycle (EOC) minus 2000 MWD /t, within one hour determine that MCPR, as a function of core flow, is shown in Figure equal to or greater than the MCPR limit times the Kg 3.2.3-1 with:
a.
If ODYN OPTION A analyses are in eff ect, the MCPR limits are listed below:
1.
MCPR for 8x8 fuel
= 1.29 2.
MCPR for 8x8R fuel = 1.26 3.
MCPR for P8x8R f uel = 1.28 -
b.
If ODYN CPTION B analyses are in effect (refer to Specification 3.2.3.2), the MCPR limits are lis'ted below:
1.
MCPR for 8x8 fuel
= 1.29 2.
MCPR for 8x8R fuel = 1.25 3.
MCPR for P8x8R fuel = 1.28
\\
BRUNSWICK - UNIT 2 3/4 2-9 Amendment No. 83 1
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued) 2.
EOC'minus 2000 MWD /t to EOC, within one hour determine that MCPR, as
~
a function of core flow, is equal to or greater than the MCPR limit times the Kg shown in Figure 3.2.3-1 with:
a.
ODYN OPTION A analyses are in effect, the MCPR limits are listed below:
1.
MCPR for 8x8 fuel
= 1.37 2.
MCPR for 8x8R fuel = 1.38
'3.
MCPR-fcr P8x8R fuel = 1.41 If ODYN OPTION B analyses are in effect (refer to Specification 3.2.3.2), the MCPR limits are listed below:
1.
MCPR for 8x8 fuel
= 1.29 2.
MCPR for 8x8R fuel = 1.26 3.
MCPR for P8x8R fuel = 1.29 b.
With MCPR, as a function of core flow, less than the applicable limit determined from Figure 3.2.3-1 initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMt1 POWER to less than 25% 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 4.2.3.1 MCPR, as a function of core flow, shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a THERMAL POWER increase of l
at least'15% of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating in a LIMITING CONTROL R0D PATTERN for MCPR.
C e
BRUNSWICK - UNIT 2 3/4 2-10 Amendment No. 83
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)
LIMITING C0 EDITION FOR OPERATION 3.2.3.2 For the OPTION B MCPR limits listed in Specification 3.2.3.1 to be used, the cycle average 20% scram time (T
) shall be less than or equal to the Option B scram time limit ( T ), where*I*
and T B
B ***
follows:
nI NT i=1 gg
, where:
T,y,
=
tt N
[
.'N.
i=1 i= Surveillance test number, n = Number of surveillaitee tests performed to date in the cycle (including BOC),
th Ng = Number of rods tested in the i surveill'ance test, and I = Average scram time to notch 36 for surveillance test i N
1/2 I
= n + 1.65 ( n
)
(a), where:
IB N
[
i=1 s
i = Surveillance test number r = Number of surveillance tests performed to date in the cycle (including BOC),
th Ng = Number of rods tested in the i surveillance test My = Number of rods tested at BOC, p = 0.834 seconds (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup on notch 36),
~
o = V.059 seconds
~
(standard deviation of the ab]ve statistical distribution).
APPLICABILITY: OPERATIONAL CONDITION 1, wheta THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.
l
~
e 2
(
BRUNSWICK - UNIT 2 3/4 2-11 Amendment No. 83
9 1
POWER DISTRIBUTION LIMITS l
1 LIMITING CONDITION FOR OPERATION (Cori.inued) l ACTION:
Within twelve hours af ter determining that T *** greater than T, the operating B
limit MCPRs shall be either:
Adjusted for each fuel type such that the operating limit MCPR a.
is the maximum of the non pressurization transient MCPR operating limit (f rom Table 3.2.3.2-1) or the adjusted
' pressurization transient MCPR operating limits, where the
' adjustment is made by:
~a'\\.
7 r
(M R MGRadjusted = M GRoption B +
~
option A option B T - T where: T = 1.05 seconds, control rod average scram insertion A
time limit to notch 36 per Specification 3.1.3.3, MCPRoption A = DeterminedDetermined from Table 3.2.3.2-1, MCPR r m Table 3.2.3.2-1, or
=
option B b.
The OPTION A MCPR limits listed in Specification 3.2.3.1.
SURVEILLANCE REQUIREMENTS 4.2.3.2 The vr. lues of T and T shall be determined and con, ared each time a scram time test is per$E>Ined.
be requirement for the f reque scy of scram time testing shall be identical to Specification 4.1.3.2.
s me t
BRUNSWICK - UNIT 2 3/4 2-12 Amendment ' No. 83
________________m__
7 O
A&
0
=/[jf //
kh %/
IMAGE EVALUATION
((//gf YD, q?,ffey:/
%,g%xf t/
TEST TARGET (MT-3)
Y 4 g+
l.0 l#2 DM gllLE I.I [" E l
1.8 1.25 1.4 l
1.6 150mm f
4 6"
i
%g
/ '/b
- 4;f b;,,,
'thfb si, y
hi 4*
sk 9
////gj
[f[$g,#,
% TA $
IMAGE EVALUATION
/ 77/ % T @/
TEST TARGET (MT-3)
+q,$
'%$[,h?
4?
4 1.0
'd m Ea gljDM i.i Liu m I.8 I.25 1.4 1.6 4
150mm 4
6"
- ks?
- k:;;f %
s, e
l g
TABLE 3.2.3.2-1 e
m TRANSIENT OPERATING LIMIT MCPR VALUES n
/-
m j
I TRANSIENT FUEL TYPE /
8x8 8x8R P8x8R g
U NONPRESSURI7.ATION TRANSIENTS With RPT operable (op.)
1.29 1.21 1.22 With RPT inoperable (inop.)
1.29 1.25 1.28 e
TURBINE TRIP / LOAD REJECT WITI'0UT BYPASS
.u MCPR N
MTR A
B A
B A
B 8
1.27 1.19 1.27 1.19 1.29 1.21 y
RPT (op.)
RPT (inop,.) B00 + EOC - 2000 1.25 1.08 1.26 1.08 1.28 1.09 RPT (inop.) i.JC - 2000 + EOC 1.37 1.25 1.38 1.26 1.41 1.29 FEEDWATER CONTROL FAILURE MCPRA B
A B
A B
RPT (op.)
1.19 1.16 1.19 1.16
?.19 1.16 if k
RPT (inop.) I}0C + EOC - 200d' 1.18 1.12 1.19 1.13 1.19 1.13 15 RPT (inop.) E00 - 2000 + EOC 1.18 1.12 1.18 1.12 1.19 1.13 D
1 o
5 i
- 1. f4 E
l-m
)
UNACCEI TABLE OPERAl f0N NN x
AUTOMA'~IC FLOW CONT ROL i
A w
1.-
i y
xx
\\
/
SCO P TUBE SETP( ' INT
/
Call BRATION POSI TIONED 1'
SUCI THAT y
FLOft/ X: 102.5%
//
=107.0%['
M
/
=112.0%/
i
= 117.3%
i 8
30
'! 0
.30 60 70 80 90 100 m
CORE FLOW (%)
K FACTOR g
FIGURL' 3.2.3-1
POWER DISTRIBUTION 'IMITS 3 /4. 2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The' LINEAR' HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kw/ft for 8 1 8, S X 8R, and P8 X 8R fuel assemblies.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or
. equal to 25% of RATED THERMAL POWER.
ACTION:
h..
With the LHCR of any fuel rod exceeding the above limits, initiate corrective action within 15 minutes and continue corrective action so that the LEGR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or reduce THERMAL POWER to less than 25% 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 />.
s SURVEILLANCE REOUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the applicable above limit:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Initially and at least once.per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.
z t
BRUNSWICK - UNIT 2 3/4 2-15 Amendment No. 83
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding i
temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.'
3/4.2.1 AVERAGE PLANAR LINEAR REAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the. limit specified in 10 CFR 50, Appendix K.
~.
The peak cladding temperature (PCT) following a postulatad loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution Lithin a assembly. The peak cladding temperature is calculated assuming a LHGR for thd highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR, times 1.02 is used in the heatup code along with the exposure-dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification APHGR is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6.
s The calculational procedure used to establish the APLHCR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2._1 5, and 3.2.1-6 is based on a l
loss-of-coolant accident analysis.
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.
A complete discussion of each code employed in the analysis is presented in Reference 1.
Differences in this analysis compared to previous analyses performed with Reference 1 are (1) The analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6; (2)
~
Fission product decay is computed assuming an energy release rate of 200 MeV/ Fission; (3) Pool boiling is assumed after nucleate boiling is Icst during the flow stagnation period; and (4) The effects of core spray entrainment and countercurrent flow. limitation as described in Reference 2, are included in the reflooding calculations.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
s t
BRUNSWICK - UNIT 2 B 3/4 2-1 Amendment No. 83
Bases Table B 3. 2.1-1 SIGNIFICANT INPttr PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BRUNCWICK - UNIT '2 Plant Parameters;,
I Core Thermal Power 2531 Mwt which corresponds to 105% of rated steam flow 6
i Vessel Steam Output 10.96 x 10 Lbm/h which correspond's
~'
l to 105% of rated steam flow t -
VesseY' Steam Dome Pressure 1055 psia I
Recirculation Line Break Area for Large Breaks a.
Discharge 2.4 ft (DBA);.1.9 ft2 (80% DBA) 2 b.
Suction 4.2 ft i
Number of Drilled Bundles 520 Fuel Parameters:
s.
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMIN LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER **
FUEL TYPES GE0 METRY (kw/ft)
FACTOR RATIO Reload Core 8x8 13.4 1.4 1.20 l
A more detailed list of input to each model and its source is presented l-in Section II of Referer.ce 1.
i This power level meets the Appendix K requirement of 102%.
To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e.,
1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.
l
' ~
~
l BRUNSWICK - INIT 2 B 3/4 2-2 Amendment No. 83s
'l e
POWER DISTRIBUTION LIMITS BASES 3 /4. 2.2 APRM SETPOINTS' The fuel clad. ding integrity Safety Limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.43 for 8 x 8 fuel, 2.39 for 8 x 8R fuel and 2.39 for P8 x 8R fuel. The ceram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less
.than 1.0 in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the-formula in this specification when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.4,3 for 8 x 8 fuel, 2.39 for 8 x 8R and 2.39 for P8 x 8R fuel.
This adjustment pay be acc_omplished by increasing the APRM gain and thus reducing the ~ slope and intercept point' of the flow referenced APRM high flux scram curvesby the reciprocal of the APRM gain change. The method used to determine childesign TPF shall be consistent with the method used to determine the MTPF.
3/4.2.3 MINIMUM CRITICAL POUER RATIO The required operating. limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel operational transients.(() Limit MCPR of-1.07, and an analysis of abnormal cladding integrity Safe For any abnormal operating transicat _ analysis evaluation with the initial condition of the reactor being at the steady state
' operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient, assuming an instrument trip setting as given in Specification,2.2.1.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
The limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass. This transient: yields the largest A MCPR. When added to the Safety Limit MCPR of 1.07 the required minimum operating' limit MCPR of Specification 3.2.3 is obtained. Prior to the analysis of abnormal operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculated by a GE multichannel {jgadystateflowdistributionmodelasdescribedinSection4.4 of NEDO-20360 and on core parameters shown in. Reference 3, response to Items 2 and.9.
s BRUNSWICK - UNIT 2 B 3/4-2-3 Amendment No.
83
POWER DISTRIBUTION ' LIMITS BASES MINIMCM CRITICAL POWER RATIO (Continuedf For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow.
The K factors shown in Figure 3.2.3-1 are conservative for the General f
Electric Plant operation with 8 x 8 and 8 x 8R fuel assemblies because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating l'imit MCPR bsed for the generic derivation of K.
f 4
At core therm,al power levels less than or equal to 25%, the reactor will be operating at minimum, recirculation pump aneed and the moderator void content will be very small.
For all designated cm. trol rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin.
With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum. recirculation pump speed.
The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown te be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very s
slow when there have,not been significant power or control red changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.
3.2.4 LINEAR HEAT GENERATION RATE The LHGR specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel pellet densification. is postulated. The power spike penalty specified is based on
.the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735
~
Suppleme'nt 6, and assumes a linearly increasing variation in axial g;ps' between core bottom and top, and assures with a 95% confidence that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking.
BRUNSWICK - UNIT 2 B 3/4~2-5 Amendment No. 83 A
(
jo UNITED STATES g g(g NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555
- y vvf SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
SUPPORTING AMENDMENT NO. 83 TO FACILITY LICENSE N0. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 DOCKET NO. 50-324 4
1.0 Introduc' tion
~
j y
By letter' dated July 29, 1983, the Carolina Power & Light Company (the I
licensee) submitted proposed changes to the technical specifications that would provide more conservative values of the Operating Limit Minimum l
Critical Power Ratio (0LMCPR), correct errors in control rod insertion times, and delete references to 7X7 fuel assemblies.
2.0 Discussion and Evaluation The licensee has requested changes to the Technical Specifications to correct an error detected in the reload analysis which was used to detennine the OLMCPR for Cycle 5 operation.
In support of its proposed changes to the Technical Specifications, the licensee has submitted MCPR data based on the approved General Electric relcad methods (Ref.1). The
'imiting Cycle 5 pressurization transient (the load rejection without bypass transient) was reanalyzed using the approveFaifd corrected ODYN code. The analytical results show that the operating limit MCPR should be increased by 0.02 for Optien A.
The OLMCPR should be larger than 1.20, 1.29, 1.26 and 1.28 for 7X7, 8X8, 8X8R and P8X8R types of fuel, respectively. The licensee also indicates that the rod withdrawal error MCPR should be increased by 0.03 for the 8X8R and the P8X8R fuel types to reflect the higher initial MCPR values. The exposure dependent OLMCPRs for both Option A and Option B are indicated in Technical Specification 3.2.3 of the proposed Technical Specification.
The s'taff has reviewed the Technical Specification changes requested by the licensee. We find that for the determination of the OLMCPR credit is assumed for operation of the highwater level (L8) trip and turbine bypass system.
In this regard, we have concluded that this subject should be treated as a generic issue, and we plan to handle it in accordance with our internal procedures for dealing with such issues. We have also determined, based on preliminary analysis, that the risk of operating Brunswjck Unit 2 without Technical Specifications concerning sJrveillance of'the highwater level turbine trip or turbine bypass systems until the generic issue is resolved is small. Accordingly, we find that the results of analyses are consistent with the proposed OLMCPRs and safety limit MCPR and conclude that the proposed OLMCPRs are acceptable for operation during the. remainder of Cycle 5.
\\
8312090109 831128 PDR ADOCK 05000324 P
PDR l
_______-______9
W
- The proposed changes to the Technical Specifications would also revise the control rod scram insertion time requirements listed in Specifications 3.1.3.3. and 3.1.3.4.
These time requirements were correct in the original Brunswick custom Technical Specifications where they were listed in terms of rod notch position. Since the rod notch position did not correspond precfSely to percent of rod insertion, an offset of some of the scram time limits was erroneously introduced. The change to the scram insertion time specification has been made to reflect the change from notch position to percent of in;,sertion an,d is acceptable.
I The licensee has also requested changes to the technical specifications that woul'd delete references to 7X7 type fuel assemblies since this type of f
fuel is no longer.used in the Brunswick Steam Electric Plant, Unit 2.
The proposed deletions involve technical specifications for power distribution limits on average planar heat generation rates.and average power range monitor set points. We have examined the proposed specifications and have found that no changes have been made to power distribution limits or set points and that references to 7X7 type fuel have been deleted. The proposed specifications are therefore acceptable.
3.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or. total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded thiY~the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement, or negative declaration and environmental inpact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusions We have concluded, based on the considerations. discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 Reference
~
1.
General Electric Standard Application ~for ~ Reactor Fue'1, ~NED0i24011JA-4, January, 1982.
Principal Contributor: Summer Sun D'ted: November 28, 1983 a
J