ML20039B708
| ML20039B708 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/30/1981 |
| From: | Mcelroy G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18025B692 | List: |
| References | |
| Y1003J01A30, Y1003J1A30, NUDOCS 8112230451 | |
| Download: ML20039B708 (28) | |
Text
p
=:
I mago;;;
NOVEMBER 1981 1
}
SUPPLEMENTAL RELOAD LICENSING j
SUBMITTAL FOR BROWNS FERRY j
NUCLEAR PLANT UNIT 3, RELOAD 4 (CYCLE 5) l a
!$555G2dafgg, GEN ER AL h ELECTRIC PDR
Y1003J01A30 Revision 0 Class I November 1981
\\
j SUPPLEMENTAL RELOAD LICENSING SUBMITTAL.
FOR i}
BROWNS FERRY NUCLEAR PLANT i
UNIT 3, RELOAD 4 (CYCLE 5)
I Prepared:
G. K. McElr Licensing Engineer Verified M. S. Charnley Senior Licensing Eng neer Approve :
for
. Eng 1, anager Reload Fuel Licensing Nb' BLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125 GENERAL $ ELECTRIC 1
Y1003J01A30 Rev. 0 IMPORTANT NOTICE RECARDING CONTENTS OF THIS REPORT
? LEASE CEAD CAREFULLY This report was prepared by General Electric sole 3y for the Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commissicq (USNRC) for amending TVA's operating license of the Browns Ferry Nuclear Unit 3.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear Plant Unit 3, dated June 17, 1966, and nothing contained in this document shall be construed as changing said contra.ct. The use of this information except as defined by said contract, or for any purposes other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
11
Y1003J01A30 R:v. 0 1.
PLANT-UNIQUE ITEMS (1.0)*
Lead Test Assemblies (LTAs), Analysis Results:
Appendix A Data for Section 4 provided'by Tennessee Valley Authority (TVA):
Appendix B Safety / Relief Valve Capacity:
Appendix C Rated Steam Flow:
Appendix C GETAB Analysis Initial Conditions:
Appendix C 2.
RELOAD FUEL BUNDLES (1.0, 2.11, 3.3.1 AND 4.0)( )
Cycle Fuel Designation Loaded Number Number Drilled Irradiated 8DB219-Initial Core 1
8 8
Irradiated 8DRB265L 2
208 208 Irradiated P8DRB265L 3
144 144 Irradiated P8DRB265L 4
124 124 New P8DRB299( )
5 160 160 New P8DRB284Z(
5 112 112 PCDRB283 (LTA)( )
5 4
4 New P8DRB314 (LTA)(4}
5 4
4 New TOTAL 764 764 3.
REFERENCE CORE LOADING PATTERN (3.3,1)
Nominal previous cycle core average exposure at 15.2 CWd/ST end of cycle:
Minimum previous cycle core average exposure at 14.8 GWd/ST end of cycle from cold shutdown considerations:
Assumed reload cycle core average exposure at 17.1 GWd/ST end of cycle:
Core loading pattern:
Figure 1
- ( ) refers to areas of discussion in Reference 1.
1
~
Y1003J01A30 Rev. O 4.
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 aFD 3.3.2.1.2)
See f.ppendix B for this data provided by the Tennessee Valley Authority (TVA).
5.
ST ANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (20*C, Xenon Free) 600 0.032 6.
RELOAD UNIQUE TRANSIFNT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
EOC5 Void Coefficient N/A* (c/% Rg)
-7.3/-9.2 Void Fraction (%)
39.9 Doppler Coefficient N/A (c/*F)
-0.23/-0.22 Average Fuel Temperature (.*F) 1343
. Scram Worth N/A ($)**
Scram Reactivity vs Time **
/
a
- N = Nuclear Input Data A = Used in Transient Analysis
- Generic, exposure independent values are used as given in " General Electric i
Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-1, Amendment 10, April 1981.
2 gs 1
4
Y1003J01A30 R:v. 0 7.
RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)
Peaking Factors Fuel Exposure (Local, Radial, Bundle Power Bundle Flow Initial Design (GWd/T)
Axial)
R-Factor (MWt)
(103 lb/hr) MCPR 8x8*/8x8R EOC5 1.20, 1.55, 1.40 1.05 6.52 106 1.25 P8x8R/LTA EOC5 1.20, 1.54, 1.40 1.05 6.47 108 1.27 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Transient Recategorization: No Recirculation Pump Trip:
Ye.=
Rod Withdrawal Limiter:
No Thermal Power Monitor:
Yes Measured Scram Time:
No Exposure Dependent Limits:
No 9.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Nominal ACPR Exposure Range Q/A P8x8R/
Transient (GWd/T)
(% NBR)
(%)
8x8/8x8R LTA Figure Load Rejection BOC5-EOC5 626 12 3 0.18 0.20 3
w/o Bypass Loss of 100*F BOC5-EOCS 123 122 0.13 0.13 4
Feedwater Heater Feedwater BOC5-EOC5 397 121 0.15 0.17 5
Controller Failure
- The eight 8x8 fuel bundles, loaded in the periphery of the core, were analyzed using 8x8R (retrofit) limits which are conservative.
3 I
e l
l Y1003J01A30 Rsv. 0 e
10.
LOCAL ROD WITHDRAWAL ERRCR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(5.2.1)*
Limiting Rod Pattern: Figure 6 Includes 2.2% Power Spiking Penalty: Yes Rod Block Rod Position ACPR MLHGR (kW/ft)
Reading (Feet Withdrawn) 8x8R/P8x8R 8x8R/P8x8R 104 3.5 0.13 18.0 105 3.5 0.13 18.0 106**
4.0 0.15
~18.5 107 4.5 0.17 18.6 108 5.0 0.19 18.6 109 5.5 0.21 18.6 110 6.0 0.23 18.6 11.
CYCLE MCPR VALUES (5.2)
Pressurization Events:
Exposure Event Option A Option B 8x8/
P8x8R/
8x8/
P8x8R/
BOCS to EOC5 8x8R LTA 8x8R LTA Load Rejection w/o 1.30 1.33 1.22 1.23 Bypass Feedwater Controller 1.27
~1.29 1.24 1.26 Failure l
Nonpressurization Events:
~8x8/8x8R P8x8R/LTA Loss of Feedwater Heating 1.20 1.20 Rotated Bundle Error
- /1.26 1.26/*
Rod Withdrawal Error
- /1.22
'l.22/*
- Indicates cat point selected..
- Values of ACPR obtained for 8x8R fuel were conservatively applied to the
~8x8 fuel bundles in the core's periphery.
'4
Y1003J01A30 Rtv. 0 12.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3) s1 y
Transient (psig)
(psig)
Plant Response MSIV Closure 1236 1272 Figure 7 (Flux Scram) 13.
STABILITY ANALYSIS RESULTS-(5.4)
Rod Line Analyzed: 10$% Rod Line - Natural Circulation Power Decay Ratio:
Figure 8 Reactor Core Stability Decay Ratio, x /*0:
0.79 2
/
Channel Hydrodynamic Performance Decay Ratio, x2 *0 8x8 Channel:
0.37 8x8R/P8x8R Channel:
0.29 14.
ROTATED BUNDLE ERROR RESULTS (5.5.4)*
Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial CPR Resulting CPR Resulting LHGR 1.26**
1.07 17.74 15.
CONTFOL ROD DROP ANALYSIS RESULTS (5.5.1)
Bounding Analysis Results:
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity-Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant Specific Analysis Results:
Parameter (s) not bounded: _None
- See Appendix A'for LTA Rotated Bundle Results.
- Initial MCPR-given is for P8x8R fuel and conservatively applied to 8x8R fuel bundles.
-5 i
Y1003J01A30 Rev. 0 16.
LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2) l See " Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3,"
July 1979, NEDO-24194A, as amended.
17.
REFERENCES 1.
" General Electric Boiling Water Reactor Generic Reload Fuel Application,"
NEDE-240ll-P-A, August 1979.
2.
Letter No. MFN-147-80/REE-046-80, Ronald E. Engel to J. S. Berggren, General Electric Company Licensing Topical Report NEDE-24011-P-A Generic Reload Fuel Application, Amendment 8," August 26,1980.
3.
Amendment 11 to Reference 1 (to be issued).
4.
Reference to be supplied by TVA.
6
Y1003J01A30 Rev. O s
bMMMMMMM.
BBEME4ME4BEM98Bei ee!+Ei49898 4 4 BEE +iBEBiBiss
.E4BEBEMBi!4BIBEMBiBEBEBE.
- BBBBBBBsBiHi4BEBBBEBEE4BEBEBE
- MMMMMMMBEMMMBIE4BE!1
- BEMBEBEME4Ma4MBEBi!4:4:4M
- M BIBEBIBitsBBBIBEM M BIBE M E
'::BEMMBEBiBEE4E4BiBBBi!4 4BiBi
' :BEMBEBBBBBBBi!4:43232:4SEE+EBE
':: B i M M B E M M B E E4 : 4 M 9 8 4 S i B 2 4 "BEBEBEBEME4 BEE 4 M M BEE 4BE"
""BEBEBEBEBRE+EMBEMBEBE**
l
!!4 M M M M ME4 ME4:
bb+bblbb+bb+bb+bblbb+bb 4
GB@ G7[E38E @E @ OfE E b 2
IIIIIIIIIIIIII 1 357 911131117192123252729313335373941434547493!53555759 FUEL TYPE
=
P80RB299 R4
.08219, IC*
F A
=
=
8 DRB265L, R1 G=
P8DR B2842, R4 8
=
P8DRB283 (LTA), R4 C=
P8DR B265L. R 2 H
=
PSDRB314 (LTA), R4 D=
P80R B265L, R3 1
=
8DRB265L. R1 E
=
Reference Core Loading Pattern 7
Y1003J01A30 Rev. 0 1
DELETED See Section 6 I
l l
Figure 2.
Scram Reactivity and Control Rod Drive Specifications 8-
Y1003Jbl3 30 Rev. 0
=
e 1
9 Mk, hC a' t e:: e EE 5 eMe
t a s,s m.= z
- wfo M5"
- W$rE E8M
-~ As, a 9
~nC 9
1 "c
\\
m
=
\\
M-
^
m n
8 1
W C
m
=y
=~
u N
N
/
\\
g 3
1 9
d q
1 m
O 4
M
~
Q O
P g
h f*
8 Q
C i
i 3
~
o
($151N3NCdWO3 111 A!13tGW u
Ou g
o d
P 1
L y 11 9
1 Ng8 C
e m
QB i_,dg 3
m EEEd I
i m _.
- Ji
=
SS-i
=
c=
x nc i<
=
c gja f
=aks
[
smw GEB a
w&S tws:
m x
-~m.o 9
u 9
-~-wo n
n-c U
u C
w w
s,c
==
m n
n a
h 3
o CD 4
/
9 MLj:.
g b
<='
1" s,
=
l 4
-.9
.i.
..9 8
8 8
0 8
0 8
(031thi.C 1N3383d1 9
Y1003J01A30 Rev. 0
+
=
E e.
y g
O!!
a--
==:2 EWCC oWW 3
g
-wuo wo_
e c-e23 ettf a u. c Wa M:$.
of!d oa k-Sb W2 g
g
-mmroe
-mm u
I M
\\
d to
~
~
r m
E a-u 8
8y g
x F
w mc.
e, o
a j
cs 1
d r
d 3
\\
. 2
. 2 t
G G
i A
E f
- g
.n.
N W
W
.o g
o o
i 9
($151N3N0dWO3 11tA!1303W mm O
a E
o 5
5 d
es L
1 e
=
W2B m
I l
a:d m
a io y
ua
,2 o
k.
w^ ['
Ce i
aL S-w ro' c 5
m
=w S
G rw-ga n
Esr-
= a'2E c.
u
$o I r;
dy -
~
c
" tee
/
C05
.3 h u g,w,Wj
_ g.
- g m
o-2.
.- ~ e
~- e M
M#
)
~
~
=-
u
_ ta r
8-8-
-m i
i e'
bJ-A 1.
l
\\t
. 9
.9 m
3 F
.ci
.d 8
8 8
o S
8 8
o 103108 41N'3W3dl
/
10
l 1
lll1
.lllli' d8OD>"
cg* o n
i SP (WEW T
u T
EO lTY SLLL TviT IFfF T!iI R EbE lTVV vCII SVMV iATT EL 0
TfCC 0
f W
2 2
ACffU 2
P y* S 3
ERR 7
/
I f
f RE E
i L
SEIf h
0P SF!P i
1PF1 EATTl 00C5 vS Ui V0SM
!2 45 I23 5
5
,t l
1 3
C
?
j E
2
^
5 l
1 C e E
E S r M
u
(
I l
.T
.L i
01 1I0M a
T F 18 n"
r
}Z 5
r e
l lo
-5 W
i 8
C tno 3
r 2
e 1
t n
.O
. 7
.O a
5' 5
5 I
0.
2 2
w i
2 2
d 1
e E$f g NW e
F o
t TR e
I X
M s
U S
n L
u u
o F
P p
E W T
SWO R
- OL s
Em ~
EfNO FLFH e
H(
1 R
x 5
RMAL u~
- AFF t
bCiI 0
HET 0
y
\\
\\
N5 E
2 n
fftL 2
C1?F
/
fiL a
NHNN l
ET 0U l
nSJI lLNM Pl P
I
)
LEI l
} 1 ' s3 ESBD 1
ui:
VSR rVDD EEUE wRCE LVTT I2345 123T5 5
5 5
1 l
1 1
e l
C CE r
E 1
Q 1
u 5
2 S u g
(
1 i
E EM F M
I
.I
.T T
0 0
1 p
1 L
u r
Z u
- 5
-5 y-8 M
1
~
- 7
.D 0
0 m
0 0
0 0.
0 0
5 5
5 0
5 1
I 1
1 g b Eu$'.-
p ll
.1ll
,ll
Y1003.131A30 Rev. 0 t,
u 2
6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 6
6 6
6 55 38 36 38 51 6
0 6
6 0
6 47 38 34 34 38 34 34 38 43 6
0 10 0
0 10 0
6 39 34 44 40 36 40 44 34 35 4
10 0
12 12 0
10 4
31 34 40 44 40 44 40 34 27 4
10 0
12 12 0
10 4
23 34 44 40 36 40 44 34 19 6
0 10 0
0 10 0
6 l
15 38 34 34 38 34 34 38 l
11 6
0 6
6 0
6 7
38 36 38 3
6 6
6 6
NOTES:
1.
ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC (FITLL CORE SHOWN).
I 2.
NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48.
BLANK IS A WITHDRAWN ROD.
3.
ERROR ROD IS (18,35).
Figure 6.
Limiting RWE Rod Pattern 12
l l11 l11!1ll ll 1
b8Uo5$
n0*
l i
SP 5~
Y.
IWHW T
T EO0O IY T
SLtL YVTI IFFF TIII R
ITVV EEE VCII SVVV u 4 IAT T
4 ELL L
TECG FfF W
a 6 EFF F
6 F
PVV EE EARR LYF5 RE ETE5 LN SEIA 0PA R
E A qi SF r
1PRT
/
VSfB 1234S6 8
0 234 u 4) 4 C
2 ES l
I CE E
S M
(
q I
)
2.T 2.E m
M a
3 1I' t
\\
3I r
T c
S 32 xu l
6 I
6 F
a s
l
(
1 e
u u
r i
us s
3 o
12 2
1 l
0 4
C 0
0 0.
D g
0 1
0 0
0 V
3 2
1 I
S
=h.
b 0yp M
o t
TR e
s X
h n
UL o
y F
P p
E W T
SWO s
AW OL e
EO FLFW R
HL EFM O
X F
LCi 4
HET u
4 n
FFE 1
6 df5F 6
a FL NS E
NRN EI l
OUI ILN R
3 P
RS LEIW T
E ES8 U
UER VSR L
EVO EEU t
' 8 NAC LVT F
7 I2345 8
I234S 4l e
IS SI i
4l C
E c
r E
E ug t
E E
F I
I 2.T 2.T 3
'V 9l-3 1
f 6
V 6
3^
1 l
.0
.7 0
0 0.
M 0
0 0
0 5
0 0
0 1
1 2
1 g @ l6
$ h -
C 1
l' li l1 llll l
Y1003J0Li30 Rev. 0 l
i I
.i A NA1 URAL CIRCU.ATION 8 10:
PERCENT R'30 t !*4 C UL'. PERFORMAT.E LIM!i t
1.00 A
ox
.75 m
X d
I<
.50
\\
s
/
8 i
.25 H
0.00
~
- 0. 0
,20.0 40.0 G 0. 0 80.0 - 100.0 120.0 PERCENT POWER i
Figure 8.
Reactor Core Decay Ratio versic.s Power 14
Y1003J01A30 Rev, 0 0
-6 C
D A
-10 G
k 83 d
D
-15 a
i
^
w 2
u.
2
-20 8
5 g
A CALCULATED VALUE - COLD 8
8 CALCULATED VALUE - HS8
_3 C BOUND VAL 280 cal /g COLD D BOUND VAL 280 cas/g HS8
-30 8
I I
I I
I
-35 O
500 1000 1500 2000 2500 3000 FUEL TEMPERATURE PC) s Figure 9.
Doppler Reactivity Coefficient Comparison for RDA 15
Y1003J01A30 Rev. 0 20.0 A ACCICENT FUNCTION B BOUNDING VALUE 280 CAL /C n
m Aa 15.0 u
C 4
M g
.gCC C
x 9
v u
10.0 aw Q
H W
v 5.0 J
<w cc
- 0. 0 0.0 5.0 10.0 15.0 20.0 ROD POSITION. FEET OUT Figure 10.
Accident Reactivity Shape Function at 20*C 16
Y1003J01A30 Rev. 0 20.0 A ACCIDENT FUNCTION B BOUNDING VAUJE 280 CAL /G
.E 15.0 a
- 2 5
,#=
O
.c a
v w
10.0
[
C 7
s 5.0
/
U l
w
/
ce O.0 0.0 5.0 10.0 15.0 20.0 ROD POSITION, FEET GUT Figure 11.
Accident Reactivity Shape Function at 286*C 17
Y1003J01A30 Rev. O 40.O A SCF AM FUNCTIC 4 B B0llNDING VALU!: 280 CAL'0 m
E a
2 30.0 m
W 3
0
.C 9
w a
I 20.0 t
U
..e U
j
>=
/
10.0 C<
U 0.0 2
- 0. 0
- 1. 0 2.0
- 3. 0 4.0
- 5. 0
- 6. 0 ELAP5ED TIME, SECONDS Figure 12.
Scram Reactivity Function at 2')*C 18
Y1003J01A30 Rev. 'O 60.O A SCF AM FUNCTIO'd B BOL NQlNG VALul: 280 CA 0 50.0 aec m
m 3o j 40.0 u
>=
d 30.0
/
o 2
wz w
20.0 C
E t3
$ 10.0 f
a:
- 0. 0
- 0. 0
- 1. 0 2.0
- 3. 0 4.0
- 5. 0
- 6. 0 ELAPSED TIME, SECONDS Figure 13.
Scram Reactivity Function at 286*C 19/20 F
Y1003J01A30 Rev. O APPENDIX A LEAD TEST ASSEMBLIES FOR BROWNS FERRY UNIT 3, RELOAD 4 1.
DISCUSSION A discussion of the LTA geometry and impact on licensing issues is referenced in a letter from L. M. Mills of the Tennessee Valley Authority (TVA).to H. R.
Denton of the Nuclear Regulatory Commission (NRC) dated November 4,1981.
~
4
~
2.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
FOR LTA*
'1 Rod Block Rod Position ACPR
~ MLHCR_(kW/ft)
Reading (Feet Withdrawn)
~
LTA 104 3.5 0.13 18.3 105 4.0 0.15
'18.8 106**
4.0 0.15 18'.8 107 4.5 0.18 w: 19.0 108 5.5 0.24 19.1 109 6.5 0.27 19.1 110 7.0 0.29 19.1 3.
ROTATED BUNDLE ERROR RESULTS FOR LTA***
Fuel Designation: P8DRB314(LTA)
Initial MCPR Resulting MCPR Resulting LHGd 1.28 1.07 s
- Limiting rod pattern for the LTA has the error rod located at (26,35).
- Indicates setpoint selected.
/
- The rotated bundle error for the LTA does not affect the limits due to special loading surveillance requirements to preclude a misoriented bundle.
21/22
l
=
p x
Y1003J01A30 Rev. 0 APPENDIX B
~
SHUTDOWN MARGIN DETEMINATION
~-
B.1 BASES'
~
The reference loading pattern, documented.in item 3 of this supplemental reload submittal,' is the bacis for all] reload licensing and operational plan-ning and is comprised of tha fuel bun'dles designated in item 2 of this supple-l'n turn is bas'd en the best possible prediction of the mental submitical. It e
core condition at the,end of the present cycle and on the desired core energy capability for th.) reload cycle.
I t.
is designed with the intent that it will represent, as' closely as possible, the actual core loading pattern.
a, B'. 2 CORE CHARACTERISTICS ~
~. s This reference cora is analyzed in detail to ensure that adequate cold shut-down margin eripts. This section discusses the results of core calculations for shutdown aargin.
B.2.1 Core Ef f ective Multiplication and Control Rod Worth Core effective multiplication pnd control rod worths were calculated using the TVA B$T siru:ator code (References B-1. B-2) in conjunction with the TVA lattice physics data generation code (References B-2, B-3) to determine the core reactivity with'all rods withdrawn and with all rods inserted. A tabu-lation of the results is provided in Table B.l.
These three eigenvalues (effective multiplication of the core, uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 5 core average exposure corrssponding to the minimum expected end-of-cycle 4 core average exposure. 'The core was assumed to be in a xenon-free condition.
Cold k,gg was calculated with the strongest control rod out at various expo-sures through the cycle. The value R is the difference between the strongest s
rod out k,gf BOC and the maximum calculated strongest rod out k,ff at any at
(
23
e Y1003J01A30 Rav. 0 exposure point. The strongest rod out k at any exposure point 13 equal to df or less than:
~
SRO k
= (Fully controlled k,ff) BOC + (Strongest Rod Worth) BOC + R B.2.2 Reactor Shutdown Margin Technical Specifications require that the refueled core must be capable of being made subcritical with 0.38 percent ak margin in the most reactive condi-tion throughout the subsequent operating cycle with the most reactive control rod in its full out position and all other rods fully inserted.
The shutdown margin is determined by using the BWR simulator code to calculate the core multiplication at selected exposure points with the strongest rod fully withdrawn. The shutdown margin for the reloaded core is obtained by subtracting the k, given in Table A.1 from the critical k f 1.0, df resulting in a calculated cold shutdown margin of 1.1 percent ak.
Table B.1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ROD WORTHS - NO VOIDS, NO XENON,~20*C Uncontrolled, K 1.122 Fully Controlled, K 0.958 Strongest Control Rod Out, K, 0.989 R. Maximum Increase in Cold Core Reactivity 0.000 h
24
ll Y1003J01A30 Rav. O REFERENCES-B-l'.
S. L. Forkner, G. H. Meriwether, and T. D.'Beu,'"Three-Dimensional LWR Core Simulation Methods," TVA-TR78-03A, 1978.
B-2.
B. L. Darnell,-T. D. Beu,:and G. W. Perry, " Methods for the Lattice Physics Analysis of LWR's," TVA-TR78-02A, 1978.
B-3.
" Verification of TVA Steady-State BWR Physics Methods," TVA-TR-79-OlA,.
1979.
)
25/26
Y1003J01A30 Rev. O APPENDIX C Safety / Relief Valve Capacity at Setpoint (No./%)
(13/83.77); (12/77.33)*
6 Rated Steam Flow 14.09 x 10 lb/hr GETAB Analysis Initial Conditions:
Reactor Pressure 1035 psia Inlet Enthalpy 521.5 Btu /lb
{
- Analysis assumes one safety / relief valve out of service.
27/28 (FINAL)
...._m...
._.m 0
0
.I i
t I
i.
l l
u i
GEN ER AL h ELECTRIC l
l
-