ML19320D007
| ML19320D007 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/14/1980 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-80-17, NUDOCS 8007180424 | |
| Download: ML19320D007 (15) | |
Text
- ?
p V<P*%,
J Consumers 4
Power
%s'y @ / Company
,7 c, g tl General officee: 212 West Michagen Avenue. Jacuson, Micmigen 49201. Area Coce str 788-osso h
July 1h, 1980 Mr James G Keppler Office of Inspection c.s 2nforce=ent Region III US Nuclear Regulatory Co= mission 799 Roosevelt Road Glen,Ellyr., IL 60137 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT FINIT - RESPONSE TO IE BULL *dTIN NO 80 FAILURE OF 76 0F 185 CONTROL RODS TO FULLY INSERT DURING SCRAM AT A BWR IE Bulletin No 80-17, dated July 3, 1980 required for plants without ATWS related RPT, an analysis of the net safety of derating such that, in the event of an ATWS, calculated peak pressures do not exceed the service Level "C" limit (~/1500 psig) by ttking into consideration the heat re=cval capability of safety valves, isola-tion condenser, by, ss to the =ain condenser and other available heat removal sys-te=s.
This analysie was to be completed within 10 days from the date of the bulle-tin and submitted under the provisions of 10 CFR 50.5h(f).
Consumers Power Company has concluded that the Big Rock Point primary system vill not exceed design pressure in the event of an ATWS; therefore, changes in plant operating parameters such as power level vill show no net increase in the margin of safety with respect to the code pressure limit. The attached analysis provides the basis for Consumers Power Company's conclusion.
David P hoffman (Signed)
David P Hoffman Nuclear Licensing Administrator
(
CC Director, Office of Nuclear.eactor Regulation
%\\
Director, Office of Inspection and Enforcement NRC Resident Inspe. tor Big Rock Point 3 pages gg O
CONSUMERS POWER CCMPAfiY Big Rock Point Plant IE Bulletin 60-17 Docket 50-155 License DPR-6 At the request of the Commission and pursuant to the Atomic Energy Act of 195h and the Energy Reorganization Act of 197h, as amended, and the Commission's Rules and Regulations thereunder, Consumers Power Company submits our response to IE Bulletin 80-17, dated July 3,1980 entitled, " Failure of 76 of 185 Control Rods to Pully Insert During Ecram at a BWR".
Consumere Power Company's response is dated July 14, 1980.
CONSUMERS POWER COMPANY By R B DeWitt (Signed)
R B DeWitt, Vice President Cvorn and subscribed to before me this lhth day of July 1980.
Dorothy H Bartkus (Signed)
(SEAL)
Dorothy H Barthus, Notary Public Jackson County, Michigan My commission expires March 26, 1983.
9
ANALYSIS FOR ITEM 7 CF IE BUL'.ETIN NO 80-17_
Big Rock Point is a non-Jet pump BWR in which steam separation takes place external to the reactor vessel. The pri=ary coolant syste= has a design pressure of 1700 psig and a code allovable pressure (10% over design pressure) of 1870 psig. The primary steam drum is equipped with six (6) spring loaded safety valves each of which has a flow espacity at its setpoint pressure (a-1500 psia) of approximately 33% of rated steam flow.
The safety valves were sized and set based specifically on a turbine trip without bypass event in which the reactor scram system was assumed to fail.
No other event could cause such a rapid increase in system pressure and reactor power as this event. The analysis of this event for the original plant design is discussed in Chapter 12 (page 28) of the Big Rock Point FHSR and in Refer-ence 1.
This analysis shows that primary coolant system design pressure vill not bd exceeded even in this extreme case. Analysis of the containment response to such an event is presented in Reference 2.
An analysis has been conducted using the Retran computer code (Reference 3) to verify that the primary coolant system transient results presented previously are still applicable to the plant as presently operated.
Although this anlysis has not been completely reviewed per the CPCo QA program, the results are believed to be accurate sul are presented to provide further assurance that previous conclusions remain valid.
A comparisen of important plant and core parameters for the new and previous analyses is presented by the attached Table 1.
Significant assu=ptions made in this new analysis are as follows:
1.
The e=ergency condenser was conservatively not modeled. The e=ergency condenser has a capacity to remove between 5 and 10% of rated core power.
2.
Operator action to manually trip the recirculat'.on pu=ps or actuate the liquid poison system was not considered. Either of these actions vould result in a significant reduction in core power and therefore in steam production rate.
3 End of cycle void reactivity feedback (times a 1.25 factor of conservatism) was assu=ed.
k.
For each safety valve, a constant capacity equivalent to that at 1550 psia (the setting of the firss valve - other valves are set higher by consecutive increments of 10 psi) based on the Moody critical flow model (Reference h) was assumed. The Moody model predicts critical steam flows that are approximately 2% higher than the code rating of the valves. A valve accumulation of 3% of the setpoint was also assumed.
Results of this analysis are presented on the attaened Figures (1) thru (9). Present-ed are core power (Figure 1), reactor pressure (Figure 2), steam drum pressure (Figure 3), steam drum water level (Figure h), void reactivity (Figure 5), vessel outlet plenum void fraction (Figure 6), recirculation flow one of two loops (Figure 7), vessel lower plenum coolant te=perature (Figure 8), and safety valve flow (Figure 9).
Because the safety valve closing characteristics were not explicitly modeled, Figure 9 may be used to establish the number of safety valves that may open in such an event, but should not be used to determine the rate of valve cycling. The maxi-
=um predicted valve flow rate is 3h6 lb/see which is the approximate capacity of four safety valves. Hence, there exists considerable excess safety valve capacity for providing over-pressure protection in the event of an ATWS. The maximum predicted
2 primary coolant system pressure is 1671 psia at the pump discharge at 11.8 seconds g.fter the turbine trip. This pressure is well within code limits 3 and in fact, is less than primary co. ant system design pressure.
6
3 ANALYSIS FOR ITEM 7 0F IE BULLETIN 80-17 References 1.
" Transient Analysis Consumers Power Company Big Rock Point Plant", APED h093, October, 1962.
2.
" Anticipated Transients Without Scram Study for Big Rock Point Plant", NEDE-21065, October, 1975 3
"RETRAN - A Program for One-Dimensional Tranaients Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", EPRI CCM-5, December,1978.
e h.
Moody, F. J., " Maximum F4ov Rate of a Single Comp 9nent, Two-Phase Mixture",
J Heat Transfer, 81, 13h-lh2, 1965 I
4 4
i i
t c_..
,, -,. - - +,, - -.,.
h ANALYSIS FOR ITEM 7 0F IE BULLETIN 8b-17 t
Table 1 4
4-Parameter FHSR New Analysis Core Power, (Mwt) 240 2hh.8 Reactor Pressure, (psia) 1500 1350 i
f K ", ($)
57 6.5 y
Total Safety Valve Capacity 3
I
(% of ' rated -270 lb/sec) 200 192 4
Setting of First Valve, (psLs) 1700 1550
- Void reactivity at Full Power i
s I
g I
4 4
l 1
i i
e
,w,,--
..~,-..-n.
- ~+-,--,>-, -
5 O
O i
i i
i r
Cx
- o, u
U')
O N
2 o
3 cv
- n LO C
c_
O O
cm a
w a
3 s
=
2 u
g I L.:
O a-u c,
x O
e u g
r u
=a z
c ez o
a e
w z
T
~
u o,
Z a
e
. I xna -
8 s
_s
.m
_ ~ ~
o
.e _._
l t
I i
7 tc sz cz c1 t1 EP B3M0d 'WSJN W31SAS l
C
1 6
t i
4 oe l
l l
l E
i C
o
.x U
i (n
O i
.N 2
o
- ta D
~
cn Tc.
as a
jj e
i m
o 13 s
g 2
u u
o C-Cn D
g e
o
=
+
"U i
E u
N Z
s E
e 8
x a
oe W
- Z C
I Ms u
o*
W i
O CD s
&g o
N cn O
t t
f I
CS! !
0L91 O E,S I 0151 ODI 05 0 -
LISd1 SS38d
'OAd O l_
10A e
-e
.,,,n
.---,- ~,,.-
7 o
e i
i i
1 i
E C
C e
O m
o N
a 2
w i
N w
C L
E a
5 D
4 a
g o
=a-s, Z
u E
a
=*
C-D e
a m
H W
C E
i w
Z s
E a
C1 8
' e
_ a f
j Je C
e e
O CD N
f.
O N
i 1
1 I
1 1
c951
-SCvl G091 52d c2LI St91 a
2 (ISdJ
'SS3Ma 0AS 011 10A
8 oe i
I I
I r
C o
e o
LO o
N 2
o i
cw
~
LO LO C
C-s C
o a
o g
N s-.
I 2
y u.J g
C-LO as c
e_
a E
H "w
m E
l
~,
i z
i s
'D w
M a
i D
e o
s.
[
z 1e w
u a,
e o
co N
N l
c c
N CD O
\\
1
+
l t
f I
I FG'E C2'E
- G C CL 2 PS~2 C2'3
[.L 3 3.:1 )
13A37
'XIW 011 10A ej
-ma n
O 9
e 9
1 l
c c
I I
I i
Ce o,
U LO O
N 3
e i
cy CO
~
CO C
C._
N L
o "C
O r
N U
2 u
ci M
b C
,M
-o C
r
=
,u E
p 3
(C 0
M h
a t
+
w C
l x
w l
O Cll g
w h-C 1
CO N
r-O o
w_
c4 s
i Q
O I
I i
1 7 E: '
s p-0 R
(
'JB3B OIGA B013628 l
.oi
M 0
6 1
M A
0 R
1 4
8 1
C S
0
/
W 0
l i
2 N
O 1
S I
S TC R
AR P
F Y
D B
0 I
O I
i 0
V 1
0 M
/
N 1
t W
C E
L E
P P
S T
I E
I L
R I
i 0
T T
9 E
UO M
L E
I E
N T
SSE I
V B
R i
O I
U S
6 T
E RU N
G I
A F
R l
0 E
I 1
4 R
0 6
/
7 I
i 0 0
2
/
9 0
l m "O 3
wi a?
s
.ZHoCiL O O>
Ob aO tt.
l i
11 J.
a cw l
l 1
I ra o
e a,
u O
o N
1 E
2 i
e g
e a
C C
c_
5 8
m o
~E W
o o
s 1
u 5
u c_
m c
z e
8 E
i--
L.
s r
3 u
5 z
5 a
y E
e_
8
^,
~
z u=
C 3
e u-7_,
O e
aa N
&o mN,o m
a ii i
r~~.
tot,-
essi cul Ec9i tisi LC G (33S/gl)
M0ld SSbW OE2 NDP i
l ct _
0 6
1 t
t A
0 R
i i
4 1
C S
0 E
/
R U
W 0
T 2
A i
e R
1 S
EP S
M E
A T
P E
Y VA B
0 l
i i
0 00 1
0 C
/
1 W
W C
t N
E EL P
S P
I I
R Ri i
0 E
T 3'
W E
O 1
L t
E L
I N
I S
E S
I E
6 V
R i
i O
U S
8 T
ER N
UG A
I F
R T
E i
i 0
4 R
0 6
/
7 0
e 0
1 2
/
9 0
$m pe nS
- e n r~ *
,b
..c j H
.O>C O~
J_ a >
OJQ-
^ u.
0:a u
i f
Q 09/07/60 RETRAN TURBINE TRIP W/0 BYPASS-W/0 SCRAM n
i i
i i
i i
U in u1 w
N LD J
m, m
i 2
Om
_J a uT wwcr E w
'n e
aro cn Z$
fj j i y l iurrrr r gii y 7 niur li 1 l Il 1 j l'
i
_J ca
- l y
9e a
i I
l l
d,' l 3
l I
iL_
L.
I I
.. _ l I
_i 20 4 f, 60 90 100 120 140 160 T I f1E I S f. C 1 0
FIGURE 9 - SAFETY VALVE FLOW
.