ML20083H920
| ML20083H920 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 10/31/1983 |
| From: | Charnley J, Hilf C, Verbryke P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20083H708 | List: |
| References | |
| 22A1694, 22A1694-R, 22A1694-R00, NUDOCS 8401090556 | |
| Download: ML20083H920 (20) | |
Text
. - _ _ -
-53
[0-'l L ' q T
22A1694 CLASSI OCTOBER 1983 y -,., ',
y 9eg. 4,
~h.: ;.l.
$?h' ll<:Q;._
.': -+
?._7., q ;,
+
r:
~
..sr- -;
i g ?:
- e. J,..
w GENERAL ELECTRIC BOILING WATER
$z$i)
REACTOR SUPPLEMENTAL RELOAD
$y; LICENSING SUBMITTAL FOR 49; sa PILGRIM NUCLEAR POWER STATICN l
UNIT 1, RELOAD 6 h
E ti h
K-9 GENER AL h ELECTRIC E
e=
E
22A1694 Revision 0 Class I October 1983 GENERAL ELECTRIC BOILING WATER REACTOR SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION, UNIT 1 RELGAD 6 Prepared:
b.'
c dtu dl4 -
P. A. Verbryke ('
hI/
Verified:
C. L. Hilf [/
Approved:,
/[d # of J./S. Charnley, Manap P6el Licensing NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL $ ELECTRIC 1
s 2
22A1694 ney, o A
IMPORTANT NOTICE REGARDING CONTENTS OF IHIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Boston Edison Company (BECJ) for BECo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending BECO's operating license of the Pilgrim Nuclear Power Station. The information contained in this report is believed by General Electric to be an accurate and true representation of the f acts known, obtained or provided to General Electric at the time this report was prepared.
y The only undertakings of the General Electric Company respecting information in this document are contained in the contract between Boston Edison Company 5
and General Electric Company for reload fuel fabrication for the nuclear system for Pilgrim Muclear Power Station, dated July 14, 1972, and nothing contained K
in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any
'Y' such unauthorized use, neither General Electric Company nor any of the con-tributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information
~
T-contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for
,q 7
liability or damage of any kind which may result from such use of such information.
r.
b A
y
~>
k 11 E
22A1694 Rev. 0 1.
PLANT UNIQUE ITDiS (1.0)*
Increased Core Flow Throughout Cycle:
Appendix A Bounding Control Rod Drop Accident Analysis:
Appendix B Safety / Relief Valve Low Setpoint:
Appendix C 2.
RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 AND 4.0)
Fuel Cycle Number Designation Loaded Number Drilled Irradiated 8DB219L 4
24 24 8DB219H 4
8 8
P8DRB265L 5
120 120 P8DRB282 5
64 64 P8DRB265H 6
60 60 P8DRB282 6
112 112 New P8DRB282 7
160 160 BP8DRB282 7
32 32 Total 580 580 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end 16474 mwd /ST of cycle:
Minimum previous cycle core average exposure at end 16074 mwd /ST of cycle from cold shutdown considerativus:
Assumed reload cycle core average exposure at end 18070 mwd /ST of cycle:
Core loading pattern:
Figure 1
- (
) refers to areas of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6 and NEDO-240ll-P-A-6-US.
1
22A1694 Rev. 0 4.
_C_ALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTil -
NO VOIDS, 20*C (3.3.2.1.1 AND 3. 3.2.1.2)
Minimum Shutdown Margin, BOC, k,gg Uncontrolled 1.113 Fully Controlled 0.955 Strongest Control Rod Out 0.983 R, Maximum Increase in Cold Core Reactivity with 0.007 Exposure into Cycle, Ak 5.
STANDBY LIQUID CONTROL SYSTD1 SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) pynt (20*C, Xenon Free) 700 0.048 6.
RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(REDY Events Only)
BOC7 +
6000 mwd /ST EOC7 Void Fraction (%)
35.7 35.7 Average Fuel Temperature (*F) 1150 1150 Void Coefficient N/A* (c/% Rg)
-6.43/-8.04
-5.65/-7.07 Doppler Coefficient N/A (C/*F)
-0.216/-0.205
-0.228/-0.217 Scram Worth N/A ($)**
- N = Nuclear Input Data, A = Used in Transient Analysis
- Ceneric, exposure independent values are used as given in " General Electric Application for Reactor Fuel," NEDE-240ll-P-A-6-US.
2
22A1694 Rev. 0 7.
RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETER (S.2.2) a Fuel Peaking Factors Bundle Power Bundle Flow Initial Design (Local, Radial, Axial) R-Factor (MWt)
(1000 lb/hr)
MCPR BOC7 + 6000 mwd /ST BP8x8R/
1.20 1.73 1.40 1.051 5.836 106.2 1.34 P8x8R 8x8 1.22 1.63 1.40 1.098 5.480 105.5 1.30 EOC7 BP8x8R/
1.20 1.67 1.40 1.051 5.634 107.4 1.39 P8x8R 8x8 1.22 1.56 1.40 1.098 5.246 107.1 1.37 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (S. 2.2.2)
Transient Recategorization:
No Recirculation Pump Trip:
No Rod Withdrawal Limiter:
No Thermal Power Monitor:
No Measured Scram Time:
No Number of Exposure Points:
2 5
9.
OPERATING FLEXIBILITY OPTIONS Single Loop Operation:
No Load Line Limit:
No Extended Load Line Limit:
Yes Increased Core Flow:
Yes Flow Point Analyzed:
107.5%
Feedwater Temperature Reduction:
No 3
22A1694 Rev. 0 10.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
ACPR Flux Q/A BP8x8R/
(% NBR)
(% NBR)
P8x8R 8x8 Figure Exposure: BOC7 to BOC7 +
6000 mwd /ST Load Rejection Without Bypass 527 121 0.27 0.23 2a Feedwater Controller Failure 318 121 0.22 0.20 3a Exposure: BOC7 + 6000 mwd /ST to EOC7 Load Rejection Without Bypass 580 124 6.32 0.30 2b Feedwater Controller Failure 336 124 0.28 0.26 3b Exposure: BOC7 to EOC7 Inadvertent Startup of HPCI 121 114 0.13 0.13, 4
11.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
(Generic Bounding Analysis Results)
ACPR Rod Block Reading (%)
(All Fuel Types) 104.
0.13 105.
0.16 P
106.
0.19 107.
0.22 108.
0.28 109.
0.32 110.
0.36 Setpoint Selected is:
107.
4
22A1694 Rev. 0 12.
CYCLE MCPR VALUES (S.2.2)
Nonpressurization Events:
Exposure Range: BOC7 to EOC7 BP8x8R/
P8x8R 8x8 Inadvertent Startup of HPCI 1.20*
1.20*
Fuel Loading Error 1.24 Rod Withdrawal Error 1.29 1.29 Pressurization Events:
Option A Option B BP8x8R/
BP8x8R/
P8x8R 8x8 P8x8R 8x8 Exposure Range:
BOC7 to BOC7 +
6000 mwd /ST Load Rejection Without Bypass 1.40 1.36 Feedwater Controller Failure 1.35 1.33 Exposure Range: BOC7 + 6000 mwd /ST to EOC7 Load Rejection Without Bypass 1.45 1.43 1.40 1.38 Feedwater Controller Failure 1.41 1.39 1.32 1.30 13.
OVERPRESSURIZATION (NALYSIS
SUMMARY
(S.2.3) sl v
Transient (psig)
(psig)
Plant Response MSIV Closure 1315 1330 Figure 5 (Flux Scram)
5
22A1694 Rev. 0 14.
STABILITY ANALYSIS RESULTS (S.2.4)
Rod Line Analyzed: Extrapolated Rod Block Decay Ratio:
Figure 6 Reactor Core Stability Decay Ratic, x /*0:
0.65 2
Channel Hydrodynamic Performance Deccy Ratio, x /*0 2
8x8 Channel:
0.23 BP8x8R/P8x8R Char.nel:
0.18 15.
LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR liisoriented 1.22 1.07 16.
CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
See Appendix B.
17.
LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (S.2.5.2)
See " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," August 1977, NEDO-21696, as amended.
6
22A1694 Rev. 0 y
MMMMMMM HMMMMMMMM 4
MMMMMMMMMMM 4
- MMMMMMMMMMMMM
-7
- ': M M M M M M M M M M M M M i
':MMMMMMMMMMMMM i
- MMMMMMMMMMMMM 3
- MMMMMMMMMMMMM s
- MMMMMMMMMMMMM i
- M M M M M M M M M M M M M i
MMMMMMMMMMM s
MMMMMMMMM HMMMMMM s
l IIIIIIIIIIIII e-1 3 5 7 91113151719212325272931?3353739414345474951 E=
a:?
FUEL TYPE C
w A = 8DB219L E = P8DRB265H d
B = 8DB219H F = P8DRB282 L-C = P8DRB265L G = P8DRB282
.g D = P8DRB282 H = BP8DRB282 bx Figure 1.
Reference Core Loading Pattern
-g n
=1
_k 7
E f
22A1694 Rev. 0 1 NEUTRON FLU (
l VESSEL PRESS RISE (PSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0
- cvoAss vai_eg ettu
-4.(( k.
=...
t ?>
.).
5 d 10 0. 0 200.0
' ?%.
2 50.0 100.O i.. +
, N? ;y e
- O 00,
.,, - - /
.x:
't_,
a 6.'! h[)
O. 0 2.0 4.0
- 6. 0 0.0
- 2. 0 4.O TIME (SECONDS)
TIME (SECONDS) r'J/.p.
- 1;
,..P 1+
y '[g
_x A
1 LEVELCINCH-REF-SEP-SKRT) 1 VOID R CTIVITY F-2 VESSEL STEAFLOW 2 DOPPLE.EAETIVITY 3 TURBIN,E STE LMFLOW 3, SCR A,,M RE. c_ T IV I TY 9,g,_,
- v. n_y _ e r_
f w, yv r.
' r. r_ e nu. Tr_ o r _ n_ u.
1.0 200.0 w
- 0. 0 N
100.0 s
lI Aty s.
W 5
M 3
-H
.I
- 0
. y e
L '
a n.
~
Q
-1. 0 9
y Ws
+
Tltjf.
$:kV l
-i s s. o
-2.0
=
- 0. 0
- 2. 0 4.0
- 6. 0 0.0
- 2. 0
- 4. 0 6.0 TIME (SECONDS 3 TIME (SECOND$)
Figure 2a.
Plant Response to Load Rejection Without Bypass,
~
+
100% Power, 107.5% Flow at BOC + 6000 mwd /ST 8
22A1694 Rev. 0 1bEUTRON FLU (
1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY V ALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0
- ovoaSS v* lee etcy b
-F s o s. 0 200.0 l
\\
N
/
E d
(
E 1
N 50.0 200.0 2
2
- 0. 0 0.0
,,, A--
- 0. 0 2.0 4.0
- 6. 0
- 0. 0
- 2. 3 4.0 6.0 TIME ISECONDS1 TIME (SECONDS) 1 LEVEL (INCH-REF-SEP.SKRT) 1 VO!
EACTIVITY 2 YESSEL STEA9 FLOW 2 DOPP R REACT!v!TY 3 TURBIN,E S.TEWLOW 3, SCR A, REACTIVITY mrernu Ter r
---nu 1.0 m
vn_v _ e r_ a r_ r y u, vy 200.0
=
3 200.0 0.0.._
)
..V Y
7 3,a
- 0. 0 W
\\
(
-100.0
-2.0
- 0. 0 2.0 4.0 6.0 0.0
- 2. 0
- 4. 0
- 6. 0 TIME (SECONOS)
TIME (SECONDS)
Figure 2b.
Plant Response to Load Rejection Without Bypass, 100% Power, 107.5% Flow at EOC 9
22A1694 Rev. O e
150.0 1 NEUkRON FLUX ll
! VEEEEL SSESS E*SE95!)
2 AVE SLRF ACE HE A FLU 2 S*FF Tf OLVE FLOW 3 CORE IM.ET FL0d 3 RELIEF V AL,E FLC I
15 t. 0 e cter !1E7 SM 4 BfP 'S$ V ALVE FLC f
100.0 ll 0 g
gi..0 g~,-
D g
50.3 U
%j w
M l
S t. 0 rNi 1
U C 04 'O O 4 0 ;4 b :
- e. 0
- 0. 0 20.0 40.0
- 0. 0 20.0 40.0 TIME (SECONOS)
TIME (SECONOS) 1 LEY EL C INCH-R EF-SEP-SKRT) 1 V01 ) REACTIVITY 2 VESiEL STEAMFLOW 2 DCPSLER REACT!4 TY 3 TUR31NE STEAMFLOW 3, SLRh M RE A CT i.d i tt, 1.0 vgr it eg grygy7_
159. 0
' egg wa7En rLeu 0
I O
M M2 m
~' - ' "
00 100.6,y A
-y 6
=
=
Z
?
58.0 it
-1.0 W
9 d
- 0. 0
-2.0
- 0. 0 20.0 40.0
- 0. 0 20.0 40.0
(
TIME (SECCNOS)
TIPE (SECONDS]
g 1
Figure 3a.
Plant Response to Feedwater Controller Failure, h
100% Power, 107.5% Flow at BOC + 6000 mwd /ST l
10 W....
i
m 22 169t*
Rev. 0 150.0 1 NE.,TR r+ Fw A i
1 VES BEL PRESS RISEC SI) 2 AVE! SURFACE H A FLUX 2 SAF [TY VALVE FLCd 3 CCSE lNLET FLCd u 3 REL IEF VALVE FLOW
!! 3. 3
' M::
iE' 51?
200.0 I
m g
S g ~g t o s. 0 b
g 50.0 5
I:'
r
~
50.0
/r h
' ' ' n: a e :: =
c ' ;: : : :
- 0. 0
- 0. 0 20.0 49.0
- 0. 0 20.0 40.0 TIME (SECC@S)
TIME (SECONDS) 1 LEviL(INCW REF.SEP-SKRT) 1 VOI 3 REACT!v1TY 2 VES LEL STEAMFLOW 2 DCF 'LER REACTIVIT/
3 TUR31NE STEAMFLOW
- 3. SCR.t.M, REACTIVITY r r waTre rLOW 1.0 150.0 17._ erarvtuvivt _
m i
}
M 100.0,H:
2 2 C
I 32
- 0. 0 e
- .tW'v l
8 c$,v'c '
t
-=
1,
- t..
)
E Og}<.,. -
e-n, 50.0 W
-1.0 l
l U
Alt-'
- a[lY
{I-f.*-[ 37, S,i.[6','
I.' g, ';.'
k '4lm
,4 !$
O. 0
-2,0 73,dg 3
3
.e
..2
- 0. 0 20.0 40.0
- 0. 0 20.0 40.0 F,.
TIME (SECON05)
TIME (SECONOS)
?
'"<,.\\.>
g
- 's
.73
'm 1
- .%, y Figure 3b.
Plant Response to Feedwater Controller Failure,
,i.
1007. Power, 107.5% Flow at EOC
. %."4m '
,f;fi4;,.
s-A'.'.'#.
yy
22A1694 Rev. O v.
I NEUTRON FLUX 1 VESLEL PRESS RISEtPSI) 2 AVE SURFACE HEAT FLUX 2 RELLEF VALVE FLOW 3 COR E INLET FLOW 3 BYPASS VALVE FLOW 150.0
' rce r t u rv e' '
150.0
'"C: rLC" " Cr rY'
,,_K a
,lF ~
~
~
~
~,
2, 2
,G
$ 300.0 [
- 2
~
~"
~
8
./
b s.
5 W
W 50.0
$,,e 0
i
?
i c'm~
,=
~, =
- 0. 0 50.0 100.0
- 0. 0 50.0 100.0 TIME (SECONDS) flME (SECONDS) 1 LEV EL(INCA REF-SEP-SKRT)
! WO! ) REACTIVITY 2 VESiEL STEAMFLOW 2 DOP'LER REACTIVITY 3 TUR31NE STEAMFLOW 3 SCAAM REACTIVITY 150.0
' rEE:unTEo rLOu 1.0
- T0t il egarrryttv Gg b
- 3. 0
.' ~.
100.0 u-i ? -i 1?
7
'7 2
75 t
8
=
g P
50.0 W -3.0 W
[I
- 0. 0
-,.0
- 0. 0 50.0 100.0
- 0. 0 50.0 100.0 TIME (SECONDS)
TIME (SECOND$)
Figure 4.
Plant Response to Inadvertent Start Up of HPCI Pump, 100% Power, 107.5% Flow at EOC 12
3 p.,
/
'l.~.,e
. - ' e. e.
22A1694 Rev. O j
g,. (
%'. - [r s
g,.
. j'
.p,t,
- 1. '
1 NEUTRON F_UX 1 VESSEL PPESS RISE (PSI)
(
-v.
2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW
, f. v.
- 4 3 CORE INLET FLOW 3 RELIEF V7LVE FLOW 150.0 300.0
- e. ovoace un' u r_
- r. _r n_2
_9 'i -
I,..4 n _
.. j J E n
'.- s...
ca 109.O
,p
\\
260.0 4 y,' -.
n_
as
. (...? r 4 '
f g
.tp>y-a
.,c'J-M
.o 4.3.
s o. o k
aj s e0. 0 h-3 4,, 4, q
- y_,
4.--..
- 0. 0
- 0. 0
~-
- O.00.2;
- 2 21 2
^ * '
.:j.:
O. 0 5.0 C.0 5.0
.' 70-l:- J.
- y TIME (SECO*0S)
TIME (SECONDS) 4 - / A s,-
r
s !'
..",'9,
- f..
f
_Q
's.
- 3. "
dit i+'. *..i o
1 LEVEL (INC4-REF-SEP-SKRT) 1 VOID REACTIVITY e~+1; 2 VESSEL STEAMFLOW PLER REACTIVITY
.- y-3 TURBINE STEANFLOW 2g0.0
'errn.nTre ri na 1.0
, SC M REACTIVITY
' ')4.'.
vnv, era-Truirv f
y, -l _?[,; *a' hv.x..,
.6 s > As c
~
a
):
.: x..
-(
m r
100.0 s
- 0. 0..
4.
s.
8 g :.
'g
'.. ', 'b4 g
i
.s f,
I h
D[MD
~
- 8. 0
}
qi y.l.0
-f: h ;...'
g N
[
.,, wg -q I
j.%S: ';
N.
y,.,.e : -
.)
~1: x :..;,
-100.0 l
-2.0 l
i
- 0. 0 5.0 0.0
- 5. 0
'4-y TIME (SECONDS)
TIME (SECONDS)
J i..
2 j h.
.w l
$ i(-<
w,.,
' O liC*.1-
.yyy 1
f..
m:
Figure 5.
Plant Response to MSIV Closure, 100% Power, 107.5% Flow a%<R --
e
.g.
fr'h..
- -.Av 13
.en? -e '
l
22A1694 Rev. O AB ATURAL C RCULATIO 1 B1 00 PERCENT ROD LI NE CL LTIMATE $TABILITY LINE 1.00 C
c
.75 s.
m 3
1--
/
.50
/
<oo ca
.25 0.00
- 0. 0 20.0 40.0 60.0 80.0 100.0 120.l PERCENT POWER Figure 6.
Reactor Core Decay Ratio 14
22A1694 Rev. 0 APPENDIX A INCREASE CORE FLOW THROUGHOUT CYCLE The analyses performed for Cycle 7 included increased core flow through-out the cycle. There are no concerns regarding reactor internals pressure drop or flow-induced vibration as discussed in the increased core flow analysis document for the EOC-6 (NEDO-30242).
The flow-biased instrumentation for the rod block monitor should be sig-nal clipped for a setpoint of 107%, since flow rates higher than rated would otherwise result in a ACPR higher than reported for the rod withdrawal error.
15/16
22A1694 Rev. O APPENDlX B CONTROL ROD DROP ANALYSIS The cycle-specific control rod drop accident analysis has been discon-tinued for banked position withdrawal sequence (BPWS) plants based on the fact that in all cases the peak fuel enthalpy from a control rod drop accident would be much less than the 280 cal /gm limit. This change in procedures was reported and justified in Ref erence B-1.
Reference B-2 indicates this change is acceptable to the NRC.
REFERENCES B-1.
Leti.ar, R. E. Engel (GE) to D. B. Vassallo (NRC), " Control Rod Drop Accident," February 24, 1982.
B-2.
NRC Memo, L. S. Rubenstein to G. C. Lainas, " Changes in GE Analysis of the Control Rod Drop Accident f or Plant Reloads (TACS-48058),"
February 15, 1983.
17/18
22A1694 Rev. O APPElmlX C SAFETY / RELIEF VALVE LOW SETPOINT The value used in the transient analyses for the safety / relief valve low setpoint is 1126 psig.
This is not consistent with the value of 1106 psig l
reported in NEDO-240ll-P-A-6-US.
1 b
q 7
l 19/20 (FINAL)
.s
(
GEN ER AL $ ELECTRIC i
-mm i
a