ML20112H504
| ML20112H504 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 12/14/1984 |
| From: | Deng S, Jensen S, Keheley T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17139C813 | List: |
| References | |
| XN-NF-84-118, XN-NF-84-118-S01, XN-NF-84-118-S1, NUDOCS 8501170170 | |
| Download: ML20112H504 (29) | |
Text
--
XN -NF-84-ll8 SUPPLEMENT 1 SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS RECIRCULATION PUMP RUN-UP RESULTS DECEMBER 1984 RICHLAND,WA 99352 ERON NUCLEAR COMPANY,INC.
l l BRA?85KJ88ssig
XN-NF-84-118 Supplement 1 Issue Date: 12/14/84 SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS RECIRCULATION PUMP RUN-UP RESULTS Prepared by:
/w,- e O',f. t cc C2V T. H. Keheley BWR Safe Analysis Prepared by:
m, S. F. Deng ' g 8WR, Safety Analysis
[
6.m._
/.3/u/ir' Concur:
- S'. E. Jdnsen 8WR Safety Analysis ff his,11,ss Concur:
R. E. Collingyam, Manager BWR Safety Agalysis
%# /.2M g[
Concur:
A; J.,N. Morgan, Menager Customer,, & Services Engineering f
Approve:
n( k[} f'A.Cd w ns., v R. B. Stout, Manager Licensing & Safety Engineering Approve:
~
- bWh
~
u dd< ?
G. A. Soff,' Manager Fuel Engineering & Technical Services naa EKON \\ UCLEAR CO V 3A\\ Y,
\\ C.
i I
i e
NUCLEAR REGULATORY COMMISSION DISCLAIMER 1
IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was eterived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear' fabricated reload fuel or other technical services 3
provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief, The infcrmation contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of comolience with the USN RC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accaracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or 3.
Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
XN. NF. F00, 766 ki
i XN-NF-84-ll8 Supplement 1 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 EVALUATION AND RESULTS......................................
3 2.1 SINGLE PUMP EXCURSIONS.................................
3 2.2 TWO PUMP EXCURSIONS....................................
5 2.3 MCPR AUGMENTATION FOR PUMP EXCURSIONS..................
6
3.0 REFERENCES
22 im
11 XN-NF-84-118 Supplement 1 List of Tables
- 1 Table Page 2.1 Initial Conditions for Single Pump and Two 8
Pump Excursions.............................................
2.2 Results for Rapid Single Pump Excursion.....................
9 2.3 Results for Intermediate and Gradual Single Pump Excursions.....................................
10 2.4 Results for Two Pump Flow Excursion........................
11 1
I d
J i
d i
+
i 4
4 1
iii XN-NF-84-ll8 Supplement 1 List of Figures Figure Page 1.1 Reduced Flow MCPR Operating Limit...........................
2 2.1 Single Pump Excursion (Rapid)..............................
12 2.2 Single Pump Excursion (Rapid)..............................
13 2.3 Single Pump Excursion (Intermediate).......................
14 2.4 Single Pump Excursion (Intermediate).......................
15 2.5 Single Pump Excursion (Gradual)............................
16 2.6 Single Pump Excursion (Gradual)............................
17 2.7 Two Pump Excursion from 67/38% Power Flow Point............
18 2.8 Two Pump Excursion from 67/38% Power Flow Point............
19 2.9 Calculated-Power / Flow Path for Two Pump Excursion..........
20 2.10 Core Pressure and Coolant Enthalpy for Two Pump Excursion.............................................
21 4
XN-NF TA-118 Supplement l'
1.0 INTRODUCTION
AND SUfflARY Reloading fresh fuel into a nuclear reactor core requires consideration of anticipated pump run-up excursions from reduced flow and rated flow i
conditions. The Minimum Critical Power Ratio (MCPR) full flow operating limit as reported in Reference (1) is established through evaluation of anticipated operational transients which are expected to be the most limiting at rated conditions. Analysis of pump run-up events for operation at less than rated recirculation pump capacity indicates the need for an augmentation of the full flow MCPR operating limit for lower flow conditions.
This is due to the potential for large reactor power increases should an uncontrolled pump flow increase occur. The methodology used herein is consistent with that used and approved for licensing the Dresden Units (2),
The present analysis establishes the necessary reduced flow MCPR operating limit to protect the reactor fuel against boiling transition during anticipated pump run-up events from off-rated core flow conditions.
This limit is shown in Figure 1.1.
The cycle specific MCPR operating limit for the Susquahanna Units shall be the maximum of the reduced flow MCPR operating limit of Figure 1.1 or the full flow cycle specific MCPR operating limit.
llll
~
5fghc o
..gA-
?
5
,0 1
gn 0
l i
,0 l
t 1
aea h vr sreup t co i
msR t
iip i
lhC
)
m tM d
g
,0 i
e L
9 nfw t
i oo a
g t
l R
n amf i
ru t
eml a
(
pil r
oxu w
e af o
p R m l
O P
e F
C eh.
,0 6
R Mhtt g
P t
i n
C e
dm M
i h eni t
Tbal a
w o
lu l
c w
e r
F t
i d
o c
N
,0 e
e
/
c R
u d
e e
r R
o C
la 1
t 1
o 0
T e
, 6 rug i
F
,0 5
0 4
4 3
2 0
1 1
1 1
1 1
al3 E5 S gE 5E TuMFu:a i
l!l!iillI!t l!
l, l1 flI ll, fl'
3 XN-NF-84-118 Supplement 1
/
2.0 EVALUATION AND RESULTS
(
The protection of fuel thermal margin during off-rated flow operations considers flow dependent MCPR limits to assure that boiling transition conditions will be avoided during anticipated recirculation flow excursion a
transients. This is done by assuring the MCPR safety limit is not violated during recirculation pump run-up transients.
Recirculation pump run-up (excursion) events fall into two categories:
single pump excursion and two pump excursions.
I 2.1 SINGLE PUMP EXCURSIONS _
A faulty signal at the speed controllef of a recirculation coolant pump could lead to an overspeed of that unit to its maximum capacity. Due to variability of the fluid coupler response characteristics, the rate of change of pun p speed might be variable. ENC has considered three cases:
Scoop tube moves at the most rapid rate expected Scoop tube moves at an intermediate rate Scoop tube moves at a gradual rate.
The rapid rate case considers the largest expected power excursion which exceeds and is terminated by the overpower protection trip system. The gradual rate case considers the situation in which there is no significant lag between power produced in the fuel and the resulting heat transfer to the coolant.
l
4 XN-NF-84-ll8 Supplement 1 A rapid power excursion associated with a relatively rapid sweeping of voids from the core by the increasing flow involves potential axial power distribution variations and control rod motion at critical transient times.
Hence, ENC's advanced simulator code, COTRANSA(3) with 1-D core axial kinetics was used to evaluate the transient. Also, initial reactor conditions were chosen which bound the highest expected initial power to core flow ratio
( 67% of rated power and 38% of rated flow) permitted by the Susquehanna power / flow operation domain (shown in Figure 2.9).
This was designed to anticipate the most reactive.(highest void content) core condition and the greatest power excursion. Table 2.1 summarized initial conditions considered except that dome pressure was 988 psia.
Figures 2.1, 2.2, and Table 2.2 sumarize the results of a single pump being ramped with a recirculation system time constant of 5.008 seconds.(4)-
The intermediate and gradual single pump excursion events were also evaluated with COTRANSA(3). The initial conditions considere,d are also those of the rapid single pump run-up. For the intermediate case, the pump speed of a single unit used a recirculation system time constant of 6.15 seconds.- The maximum power was less than the overpower trip point. For the. gradual case, the recirculation system time constant was 10.02 seconds.
Generally, the power ascends smoothly to a new level with the clad heat flux maintaining an equivalent pace.
Table 2.3, and Figures 2.3 through 2.6, summarize these results.
Thermal margin calculations with ENC's XCOBRA thermal-hydraulic subchannel model(5) pre-dicted that the minimum CPRs during all three events were above the Susquehanna safety limit of 1.06 which has been established (l) for avoiding boiling transition.
5 XN-NF-84-118 Supplement 1 2.2 TWO PUMP EXCURSIONS A faulty signal originating in the caster flow controller may lead to an unplanned increasing demand for flow simultaneously to both recir-culation pumps. Due to the intentional design features of the control system, an error signal within the master controller would be attenuated substan-tially before being received at the individual pump speed controller. Thus, the expected pump responses to the demand from the signals originated at the master controller are relatively gradual pump speed increases. In this case, the gradual speed increase results in power ascensions charaterized by equivalent increases in clad surface heat flux.
The evaluation of this event considered the same initial conditions surrmarized in Table 2.1 and used the PTSBWR3 code.
An error signal was simulated in the master controller that increased demand at a gradual rate (less than 1% of rated speed /sec). The analysis of this event was performed to determine the maximum power increase for a given flow increase; therefore, the void reactivity was assumed 25% more reactive than expected and Doppler reactivity was assumed 10% less than expected.
Core power, vessel pressure, and recirculation flow increased smoothly such that no system trips were encountered up to about 195 seconds where the thermal power monitor trip was encountered and CPR is below the safety limit. The results are shown in Table 2.4 and Figures 2.7 and 2.8.
The results indicate that the CPR could decrease below the safety limit if the full flow reference MCPR was observed at init'ial conditions.
Thus, an augmented MCPR is needed for part flow operation to protect the two pump excursion event.
6 XN-NF-84-118 Supplement 1 2.3 MCPR AUGMENTATION FOR PUMP EXCURSIONS The above evaluation of the two recirculation pump flow ex'cursion indicated that establishment of a reduced flow MCPR limit for this event which will prevent boiling transition will also bound other events involving single pump failures.
The calculated power / flow relationship for the two pump excursion is graphically displayed in Figure 2.9 and is compared with the expected flow control line and the 108% rod line. This calculated power / flow relationship was conservatively extrapolated past the predicted scram point to the maximum allowed core flow of 105%. The evaluation of this event was performed to determine the maximum expected power if the flow was increased up to 105%. The results of the calculation also provide an expected relationship for core' pressure and inlet subcooling as shown on Figure 2.10.
Using this information, the required MCPR augmentation to prevent the safety limit from being exceeded during the pump speed increase ' was calculated. This was performed by calculating the delta CPR using XCOBRA(5) along the calculated power / flow relationship and making the extrapolated 105%
flow point to yield the MCPR safety limit.
The MCPR augmentation for even a slower pump run-up event was also evaluated along the 100% rod line up to 105% flow maximum.
Because of the inlet enthalpy lag in the two pump excursion transient case of Section 2.2, the resulting flow dependent MCPR augmentation was slightly higher for the gradual quasi-steady run-up along the 100% rod line.
The reduced flow MCPR operating limit ~ show 'in Figure 1.1 is based on the 100% rod line pump excursion. The cycle specific MCPR operating limit for the Susquehanna Units t-
7 XN-NF-84-118 Supplement I shall be the maximum of the reduced flow MCPR operating limit of Figure 1.1 or
~
the full flow MCPR operating limit.
e e
. - =.
8 XN-NF-84-ll8 Supplement 1 Table 2.1 Initial Conditions for Single Pump and Two Pump Excursions Y
Reactor Power Level (% of rated) 67 Total Core Flow (% of rated) 38 Core Pressure (psia) 994 Core Inlet Enthalpy (BTU /lbm) 491 Steam Flow Rate (Mlb/hr) 8.62 Feedwater Enthalpy (BTU /lbm) 320 Core Active Flow (Mlb/hr) 35.3 Turbine Emission Pressure (psia) 974 Initial Pump Speed (% of rated) 20 l'
l-
~N 1
7 1
a.... :
,..c.
a
.s,
.... -... n...... n
..:., +
+
..,y' e[.(..,i _,, (. l...,, '$
- _ } * ' ; 9,';. %. Y AQ.;i, N'.89E '.., t. l 'O ~ Ey",h
' W'* k 'E% "S
' ?"
.' ' % '~
.. h,l-u % $.< 5
+.1.l 'E
..,4,','.,-- A,,,', ?
,y < s.,:; : y R,*.,. *.; g k
~
y
?;.F,-'
.j. p?
, w,. e.
g f
..:..- i
...4'
' *e,,-7.,*._4.,,, x..
.?
.s.. '. :,..
9 XN-NF-84-118
- .W. V '. "{.
?./
.~.L r
~~
Supplement 1
.b'.._-. y' l.
y 4
m c-l
^
Table 2.2 Results for Rapid Single Purrp Excursion t
m-
.t f
?.s.
o
- i
. e.. 7..
n f
%-.a.
.s. ',
N.
.59[..).J r
Peak Power (% rated) 134 h ^;. W 2..
g Maximum Vessel Pressure (psia) 1016 I.1 T.M. J
= t :.
'\\
's -4 ['_ n;[
Maximum Steam Line Pressure (psia) 981
.,! N1..fd.
7
- , e.
.c e,.
.=<
r w g.,
c.
,= R M i'!?
- ?
Maximum Core Average Clad Surface 117 f.-
Heat Flux (% of Initial Core 49.
RA Average)
. %N..? i. ". -
yff ' y=f r
8:, N Q
g,'
. '%.../-
L y :g :., :;
.9 f,
.;,Y.,1.'fw.
.6 7
5 x ; e (,. x
- y?.
,x
.a
- =. -
r.
- w n
s.w.
~
/.
p.
w
.=i 1
4
.,3
- - Ag;:.i
,{, '$
, f ':
- b. tl _.h k.
rs' s.
.N
&, :," t
?
.+
,N"' 4g
... ~,
3.. h ' > '
p.?
.-~ g$ 4 5'-'
_,7
'4
'I' ?. A.e 4 e<
g 'J.
(+.
- ',5,_,,.'$ %'gyW*
^y.e
-,1
.r
~
q, b
l'
- s.*
. 'a ' +
r.
f.e.,,.; i N. -
,..,^.ir...f.
t F-k-v M*-.
.l *.. '.,-
E
-. w
.').6..
2
~
e-
.j :
e g
.'.g y.
l'
. :f.Af k
'g
,. [.. /. *i -
..':;hy 4
- g., = ( b:b-Q
.g.
3:
x <
.1
.. e
, h.' -
4-
.9jy,
U-
+
r w >.....
?:iff. e,bW.'.#
a v$ /p.' E 3.jf.'i, f f ' (*f-
'i
,.e x.
ni-
?.=
y..
>:y:_
.?;.+y
- 3.. f..
- 4. y e ;. g ; 7 : ' ;_
- ., y; y 7; 3.-3 g 9gy {;;;}Q
ei, h:S: n., ^ : %-iwv 0+
. w,4... p.r.., y,p sc m.,,. m.y~ u; g.
.y
.,,,.,.,_ 7.
- . 9 _..., ~
. ;;t,': M.
j yn,.
- 4..
?.ilx- ;T i
'X.%..
y g...,
q,
/{
10 XN-NF-84-118 M^
Supplement 1 3,.7, c.
N
.I..
5.,
9
.,,.g.
it Table 2.3 Results for Intermediate and Gradual 4."i
..i.
Single Pump Excursions
"'i. ' '::
8 c'?
c,,....
.. 'a B
Intermediate Gradual b i'],
..i :
%. 2.",.., ;
1 s
Maximum Power (% rated) 118.9 95.6
- t. 7:2.'
tm.
5,.; y -.
M.
Core Avera9e Heat Flux 82 80
~ v, e c~
1.
l<.'.. A@
(% rated after 9 sec)
.,. x,
s.
g
=,
y;:...
v k'
Active Core Flow Rate (Average, Mlb/hr)
?
i -; % '.'.
..I
/
At reported core average heat flux 43.7 42.8 EL ' { S, w.x
,,e.,,~..t..-
i..
pj;
- /
Core Inlet Enthalpy (Btu /lbm)
.}.) ;ip>.
e..
";S At reported core average' neat flux 491 491
- 9. '. 1 ;.
q.) :
f'p Core Exit Pressure (psia)
V 22:
g? :. 'g Maximum 1013 1008 4
9 ','i*
At reported core average heat flux 1008 1005
%. ~,.
r
%h u '
3
$.k..
!.-lff...
c yf'S -
N '.,1.i~
- .d..<:
g i
,4 n
.s.. '
re s
'f.:'-*..-,
_.,fg
?
.,r,
- y. ;
h..
IQ '.t c..
yy
. _ y ';..,
'\\i'.
c.
A..-f
-.{..;. '.
-.r;.'. " :
.:. ",J
(.- r,. $.
< e p.:i 'y..
Y.
"P a',~;,..
W.,
g,.
-;., - 3
/ - )..i j$
3.s 4.',.,
.? ' i N
- i p-
- 3. *q g
- r. s.,s.
.h..
5
- l >
g
'm: y __, ;. -v.
_y_;;.s.
y _:,. 3.2 :'-
7
[ o y.y.;.
p, ;; :_ y. :., ': [.: y ; : 3 ; en,[ ;;;.;- y. ;, 7.
9 11 XN-NF-84-ll8 Supplement 1
~ Table 2.4 Results for Two~ Pump Flow Excursion Maximum Power (% rated) 118 At 180 Seconds After Pump Speed Change Reactor Power (% rated) 114 Core Average Heat Flux (% rated) 109 Active ' Core Flow Rate 74.7 (Average,Mlb/hr)
-Core Inlet Enthalpy (BTU /lbm) 512 Core Exit Pressure (psia) 1044 l
1
-_i i-
I Um x2 A" vg
"~
0 5
\\
s 5
4 N
4 t
m 0
4 s
WOW LO FLF v 5
N u
IM O
Le 3
)
uL TE d
4 i
rXAT n
p ULS e
a nLU R
i uFCL n
(
n RE v
0 n
iI [ S a
3 o
uACS r.
K M
i eEEE t
\\
s C
r nHRV e
1 u
E c
_j S
x E
5 12 p
EE m
M u
P I
2 T
e lg 0
n i
2 S
g*
1 e
2 5
er 1
}
J ug
=
i F
g 0
1
- ,v s
5 t
0 g
0 i
0 0
0 0
4 2
0 0
0 0
1 1
1 8
'4 2
0 8wGE O gjn.
1!
1
1I
- i lI
- i E L.
n RE
-w 0
5
\\
N i
5 x
4 s
r i N
- e. I M
if 0
ir wL w
4 uEV eE
)
nL d
i u
p 5
sR 1
a mE R
3 eT
(
nA n
rW o
i iL t
s mBE r
1-0 uc sS 3
x eE E
vV C
p E
mu t
S P
12 5
e l
2E
/
g M
n I
i S
T 0
2
~
a 2
erug g
i F
5 1
1 5
0 1
1 5
0 0
4 0
0 0
0 0
0 6
5 4
3 2
1
~
1lll1lll li
Ese?E n
'$iag[
0 01 0
9 5
4 0
1 8
3-W 4
OW LO 2
)
FL 1
e F
w t
N o
0 a
i OM L
7 d
p u
IF F
5 e
t TE m
4 rXAT n
re ULS t
t s
t rLi a
i n
uFCL p
1 0
I n
RE w
(
s iI [ S 6
_M n
uRCS e
o tEEE t
4 C
i rHRV e
e s
3 E
r S
u 4
a~~
0 c
x 1234 s
E 5 E Ms/
M pm I
u T
P e
0 lg t
n 4
/
s i
S 7
s 3
3 0
2 3
e p
ru g
i F
0 2
at-P 0
1
/
t 0
L 0
g O
0 0
0 0
0 g
S 4
2 8
6 4
g I
1 1
a u g u.O e h u s
IIIt i
t l'
li ill ll jl(t l
,lt il
lll1l li' l
1ll 1
l l
l ll ll llll
-m
=
2b" r"2A" 0
01 0
a u
9 N
r t
N t, I 1
t (
0 n
t
- w. L
,N 8
L EV tE
)
nL t
e t
u 0
a nR i
E 7
d tT e
uR m
rW re t
.L E
n teE 0
I sS
(
yS 6
n tE o
vV C
is E
r p
S u
E c
12 0
x E
5 E p
M m
I u
T P
e 0
l t
gn 4
i S
t 4
r 0
2 3
e N
rug i
F 0
t
(
s 2
0 t
1 0
i 0
0 0
0 0
0 0
"0 7
6 5
4 3
2 illrll!
l l
x2kk g"'
mCE$R "
0 0
1 5
1 0
1 l
5 9
s 0
8 s
O 3
LF
)
o 0
l m
a 7
u X
o d
t a
eX T
o r
U S
c s
G mL
(
i F
n m
w 0
n 1
P o
ir 6
i's uA t
r
- e. E t
C u
rH e
c E
x S
E 0
p 22 3
m 5 E u
M./
1 P
rI e
T lg n
0 i
S 4
7/ ;
F 5
2 e
0 ru 3
g f#
F
[
i F
f'#
v 0
2 0
e 1
v 5
4 0
t 0
00 0
0 0
0 o
0 8
5 t
3 1
9 8
e"'
z0EE t
i
..w M
17 XN-NF-84-118 Supplement 1 o.
O m1 m
g L
~ ZH u
'o
.4, j."
kbd dC z _,
o
.4 3m
^
n uJ b
lr >E J
=
C o
s 2
.J N
O Ie
.*j o
^
a e
r un o
d Id Q
m W
Ii a
to u
in w a
T.
G H-E r
y u
.4 Q
Gc w
e.
w N
O.
h 4
C tb e>
m a
h h
O O
C o
o b
W v
M
160 1
retuinurs etu 2
HEA I FLUX i
3.
REC [RCULATION FLOW 4.
VES SEL STE0H FLOW 140 3.
ree ;waign ruov 120 O
I 100
- M ms
/
d'
[
[
3 t-5 80 t Vg
[
n.
/,
60
/
/
7 40 J-E'u 20 o*
0 9.0 40 60 80 100 120 140 160 180 200 TIME, SEC
~E Figure 2.7 Two Pump Excursion from 67/38%
Power Flow Point
19 XN-NF-84-118 Supplement 1 OoN f
C CD
,c
\\
~
s 10H n
b' O.
1
-lc" x
C
-DW h
k r
s 5
a
<a 5
g Cu
=
om dW O
'"y 9W 5E
.a cn u c.
AN O
$y C
- o d IJJ C1 -
E EL b
sa
<a
~'
1 C
h N
W Q
C h
W L
-W N,
+
ON a
ed C
C C
C C
C C
C=4 h
W to w
M N
C
I ME a
TEGo E
c 0
11 en i
,0 L
0 k
1 co l
B d
r o
R
,0 o
f 9
M h
R t
P a
A D
w e
E P
n w
e" T
o i
v%
,0 A L
l 8
F n R
/o d
h o
ri t
R es a
wr P
ou 0
P c t
0 x
ne 0
dE e
,7 W t p is am n
O l u a
uP r
L c
T F
l o aw CT
,0 E 6 RO 9
C 2
e r
,0 u
5 g
i F
,0 4
0 3
0 0
0 0
0 0
4 2
0 8
6 4
1 1
1 M
l1 l
2l.
I I
- l 5
gh g5 5
0 5
0 5
0 1
1 0
0 9
9 5
5 5
5 4
4 3
3 1
,0 2
1 ro f
y 0
p
,1 l
1 a
h t
y n
p E
la t
h n
t a
n l
E o
0 o
,0 C
)
d 1
n e
do t
ni a
as R
r e
eu r
rc
(
u ux s
sE r
s s
e e
,0 w
ep r
9 rm o
P Pu P
P e
4
- o ow CT 0
0 1
i 8 2
erug i
F 0
i 7
0
~
6 0
0 0
0 0
0 0
0 0
0 0
7 6
5 4
3 2
1 0
9 8
7 0
0 0
0 0
0 0
0 9
9 9
1 1
1 1
1 1
1 1
2Da.~ ai$u" E8 ll
22 XN-NF-84-ll8 Supplement 1
3.0 REFERENCES
1.
'S.E. Jensen "Susquehanna Unit 1 Cycle 2 Reload Analysis," XN-NF-84-116, Exxon Nuclear Company, November 1984.
2.
"Dresden Unit 3 Analysis for Reduced Flow Operation," XN-NF-81-84(P),
Exxon Nuclear Company, January 1982.
3.
G.C. Cooke and R.H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling. Water Reactors," XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, November'1981.
4.
- Transient Safety Analysis Design Report, Susquehanna Units 1 and 2, GEZ-7127.
5.
T. Patton, "XCOBRA Code User's Manual," XN-NF-CC-43, Revision 1, Exxon Nuclear Company, January 1980.
I 1
E
i 1
XN-NF-84-118 Supplement 1 Issue Date: 12/14/84 j
SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS-RECIRCULATION PUMP RUN-UP ANALYSIS 1
DISTRIBUTION
-R. E. Collingham J. C. Chandler S. F. Deng J. G. Ingham S. E. Jensen T. H.~Keheley L. A. Nielsen H.'G..Shaw H. G. Shaw/PP&L (60)
Document Control (5) 6 t-d
=
.m