ML20010E964
| ML20010E964 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 08/25/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-09, TASK-15-14, TASK-15-9, TASK-RR LAC-7756, NUDOCS 8109090203 | |
| Download: ML20010E964 (10) | |
Text
'
w D
DA/RYLAND h
[k COOPERATlVE po box 8 7 2eis EAST AV EOUTH = LA CROSSE. WISCONSIN 54601 l
(608) 788 4 000 August 25, 1981 In reply, please refer to LAC-7756 i
l DOCKET NO. 50-409
.U.
S. Nuclear Regulatory Commission
<D ATTN:
Mr. Darrell G.
Eisenhut, Director oj Division of Licensing
'/
j Office of Nuclear Reactor Regulation N
.1 Division of Operating Reactors Washington, D. C.
20555 E
D 4I g
[s
%y /'y?
SUBJEC"':
DAIRYLAND POWER COOPERATIVE s.
LA' CROSSE BOILING WATER REACTOR (LACBWR f
DPR-45\\'y PROVlSIONAL OPERATING LICENSE NO.
SEP TOPIC XV-9, STARTUP OF RECIRCULATION
'N, g i g \\,/
LOOP AT INCORRECT TEMPERATURE AND SEP TOPIC XV-14, INADVERTENT OPERATION OF ECCS
REFERENCE:
(1) r>pC Letter, LAC-7387, Linder to Eisenhut i
Dated February 27, 1981.
Gentlemen:
Enclosed find Safety Evaluation Reports (SER's) for SEP Topic XV-9, Startup of Recirculation Loop at Incorrect Temperature and SEP Topic XV-14, Inadvertent Operation of ECCS which has been prepared for the La Crosse Boiling Water Reactor.
Our letter, Reference 1, identified topics for DPC to submit for NRC evaluation, j
If there are any questions regarding these SEP Topics, please contact uc.
l l
Very truly yours, DAIRYLAND POWER COOPERATIVE Y:M<IU wh' M Frank Linder, General Manager FL: CWA:eme cc:
J. G.
Keppler, Reg. Dir., NRC-DRO III I
NRC Resident Inspectors s
.f1109090203 810835 PDR ADUCK 05000409
.P PDR
DISTRIBUTION FOR LAC-7756:
SRC Trommel Shimshak Parkyn Towsley Angle Boyd/Kelley Files:
SEP TLNB, A-ll, S-23 T5c, Rdg.
I t-
LA CROSSE BOILING WATER REACTOR SYSTEMATIC EVALUATION PROGRAM SAFETY EVALUATION REPORT TOPIC XV-9 STARTUP OF RECIRCULATION LOOP AT INCORRECT TEMPERATURE INTRODUCTION The safety objective of this topic is to verify that the plant responds in such a way that the criteria regarding fuel damage and system pressure are met (i.e., no more than a small fraction of the fuel rods fail, that radiological consequences are small fraction of 10 CFR Part 100 guidelines, and that the system pressure is limited in order to protect the reactor coolant pressure boundary from overpressurization).
EVALUATION Isolated Loop Start-Up (Hot)
This transient was analyzed in detail in Reference 1.
For this accident, it is assumed that only one recirculation loop is l
operating and that the second, which is 25*F colder than the operating loop, is started.
This results in an increased flow through the core and a subsequent increase in power.
Starting Conditions and Assumptions (See Reference 1)
LACBWR has two forced circulation pumping loops that connect to the common inlet and outlet manifolds outside the reactor vessel.
Normally, a second forced circulation pumping loop is not put into operation when the reactor is operating at a significant power level.
Interlocks have been provided to prevent a pump from starting unless its associated discharge valve is closed and to prevent the discharge valve from opening unless the temperature differ-ential between the two recirculation loops is 110*F. 1 L
m
y L
This analysis analyzes the results of establishing ful' " low in a second recirculation loop when the reactor is at potu The starting conditions and assumptions for this transient have been chosen.such that a conservative estimata of the. transient results.
The assumptions are as follows:
(1)
The reactor is initially operating at 52% of rated power.
(2)
No credit has been taken for a reactor scram.
(3)
Recirculation flow increases from 15,000 gpm to 30,000 gpm in 20 seconds, (4)
The isolated loop is 25'F colder than the operating loop.
o L
(5)
The reactor is operating at the end of a fuel cycle.
1 l
[
Results and Concequences Figures 1, 2 and 3 present curves of reactor power, recirculation flow, and turbine steam flow, respectively, versus time.
Power l
increases until at approximately 18 seconds water which entered the isolated loop at the start of the transient reaches the core.
This causes a momentary decrease in reactor power.
Power attains a final steady state level of 96%.
No damage to the core will result from this transient as the CPR stays above 1.32 at all times.
Isolated Loop-Startup (Cold)
For this accident, it is assumed that only one recirculation loop is operating and that a second which is cold is started.
This results in a small flow of cold water into the core by means of
'the bypass line around the recirculation line isolation valve.
Interlocks prevent opening of the isolation valve when the temper-ature of the water in the recirculation loop is not close to that in the reactor.
The decrease in the reactor water inlet tempera-ture results in a power rise.
The consequences of this transient are much less severe than for other moderator temocrature decrease transients which have been analyzed in detail and'found to have no adverse impact on fuel integrity or plant safety (e.g. increase in t
l Feedwater Flow - Reference 1).
2-tJ
]
L -
[
lLi l
l l
1 C
C 0
7 O
C O
'G D
C
.0 5
)
T)
O e m
I I
( i T
P tC U s S
- u
.O
-N 0
T s
,4N R r E
C A e L
T v S
S r
P e N
D O w I
C Oo
/
,/
.E LP
,0 M
3 D r I
E o T
Tt A c L a O e C
SR 0
I
(<
,0 2
1 iI, O
E C
RU
.0 G
\\'
l 1
I
/
F l
~
//
G C
cD cO -
On.o o,c.. o a m '3 av-
)
e 0
cLuoC cd
" r *~ '). C nut o c.
i r,
i u1 4
<!i w
0 O
0 0
7 0
0
.0 6
0
- 0
)
.0
) e 5
Tm Oi T I
I(
0 P s 0
U v
.S D
.0 Tw N
R o 1
iO Al C
TF E
S S
n P o N
0 Oi O
I 0
Ot L a
.E
.0 l
N 3
D u I
E c I
T r Ai L c O e 0
SR 0
I(
.0 2
2 0
E 0
RU
.0 G
I 1
F 0
0 OO.n J OO.rn Oa OW oO.e Oo.o 0 mo.n,t n ooe N0,g v't i L
o,.
4
> a J. L E J.'3 $ ~ *J nuc "otC L
c n
J 2
h T
O I
" i
!1
o o
n b
O O
.ow oa
.o
^
O
^o P s aa k
=8 v
l
\\
o aw o
g I
Tde ea
-r :-
c: o O
t H%
i o
w v3 W
E
- c. n o ~-
oe o
oa m
aw U
/
'sE co m=
~
E-<
rA
<::.o A k oa 1
o to 8 l
Q M V
.o N
M o
N o
mo
.o o
s o
O CC'Ost 00'O$
03'0!'I 00 Gil 00'001 CO *Cd (oes/q1) ;,073 Wegig grnc;gni p O I
y l
REFERENCES:
t-i.-
j.
.' " Response to Question 4 -' Transient Analysis of LACBWR Reload l-
' Fuel", prepared by. Nuclear Energy Services, Inc., NES Document
[.
No. 81A0025, February 18, 1977.
t t
t I
I l
3 I
l' L l i
~ _ :4 i
I di 5; i
LA CROSSE BOILING WATER REACTOR SYSTEMATIC-EVALUATION PROGRAM SAFETY EVALUATION REPORT TOPIC XV-14 INADVERTENT OPERATION OF ECCS INTRODUCTION The safety objective of this topic is to assure that water added to the RCPB does not cause transients that exceed RCPB pressure limits or result in unacceptable fuel damage.
No activity is released during the transient but the transient may subsequently result in increased radioactivity in gaseous releases during normal operation.
EVALUATION An accident involving inadvertent operation of the ECCS is similar to the increase in feedwater flow accident (Section SV-1).
Since the maximum flow from the ECCS-(100 gpm) is lower by an order of magnitude than the flowrate considered in the uncontrolled feedwater addition transient analysis (Reference 1).
l The reactor boundary overpressure protection systems are designed in cccordance with the ASME Boiler and Pressure Nuclear Code Cases and j
Section VIII.
The reactor pressure boundary is protected by three main l
steam 3" x 6" spring-loaded safety valves with set-points of 1390 psig l
and 1426 psig.
Therefore, continued inadvertent operation of the ECCS would not cause unacceptable fuel damage since the set-point relief pressure is significantly less than +'.at required to collapse the free-ctanding fuel cladding.
This accioent poses no threat to plant safety.
i i l
r.
~r...
.]
REFERENCES:
--" Response to Question 4~'
Transient' Analysis of LACBWR Reload Fuel", prepared by. Nuclear Energy Services, Inc., NES Document No. 81A0025, February 18, 1977.
.g
..f '.
. g..
.,