ML20090F979
| ML20090F979 | |
| Person / Time | |
|---|---|
| Site: | Crane, Yellow Creek  |
| Issue date: | 10/14/1977 |
| From: | Lumpkin R ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Patterson D TENNESSEE VALLEY AUTHORITY |
| References | |
| TASK-*, TASK-06, TASK-6, TASK-GB B&W-0543, B&W-543, TD-CE-900, NUDOCS 8307080711 | |
| Download: ML20090F979 (6) | |
Text
,_
. _. o
.~
Y Tct. 203/689191 I
~
'.. *. C E Power Systems Tclex: 9o297 Combustion Er.gineering. Inc.
1000 Prdspect iI 11 Road Windsor. Cen.,ect cut 06005 N8M2-50-0]O2
. "72 P OW ER
~
E SYSTEMS yd _gm1T [% IC210D(1.
October 14, 1977 g y. i
~
TD-CE-900 tr i!
j/fMM.L003 q
E:,
q..
F
,~~
, i.
is ;t :
Tennessee Valley Authority C
i A00 Commerce Avenue Knoxville, Tennessee 37902 g
- 5. '
(-.'. ::
Attention: Mr. D. R. Patterson, W'0C125 G'
Chief Mechanical Engineer m-i- *I
'k Gentlemen:
Subject:
Nuclear Steam Supply Systems
[t. -
Proposed Yellow Creek Plant TVA Order No. 60-84840 p g' C-E Contracts 14074 and 14174 VERY SMALL BREAK LOCA
Reference:
A) TVA letter C-655, dated July 22, 1977.
The attachment to this letter responds to the TVA concerns, expressed in Re.ference A, regarding a very small break LOCA.
ft should b:: noted that the small break post-LOCA long term cc. ling analy-
~
ses and procedurcs discussed herein are based on mi:thodologies undergoing current review by the MRC.
It is expected that NRC review of these topi-cal reports will not be ccmpleted before April 1,1978.
k i
Very truly yours, I
t i
4 R. L. Lumpkin,
{
%ProjectManager RLL/ TEM:dfm T-PPE-lGO cc:
M. J. Epprocht, C-E W. Wade, TVA OCT i 81977 i'
i.
i"dSTER FILE m^,mn?A rc'>.?l o"^Tsj-3g%
uae;^-
r% po:a:rT
_i m.+.
t.,
4 i
i 1)
%.g.. _g.e.r,4..,n.. - -_-'j F.T T____wai r i.
.?-
3
{
h.
1 062 L -- - '
p.tr i~~
f hI'Y-
- T. j~RW.h.
- .)... N t-8307000711 771034 PDR ADOCK 05000289 HOL k
~
r e
=-
T0-CE-900 C-E Resconse to the TVA Concerns Recardino Very Small Break LOCA Assumptions:
To assess the impact of a non-condensible gas on steam generator heat 1.
transfer during a small break LOCA, the presence of hydregen in the Reactor Calculations performed with the Coolant System (RCS) was inve'stigated.
assumption that ene percent of the core wide clad reacts prior to the onset of primary steam condensation and all of the gas evolved is as-sumed to collect in the steam generators, resulted in a reduction in Reaction rates typical of the heat transfer rate of only six percent.
small break LOCA response demonstrate reaction percentages much less than 0.1% prior to the start of condensation.
Consideration of SIT N2 injection is not necessary for these. small breaks because the SIT's do Since the condensation coefficient not inject during these transients.
utilized in the small break LOCA transient calculations are based on conservatively low values, (50 percent louer than best estimate values) the small reductions in heat transfer rates, as a result of non-condensib' gases in the system, are more than bounded by the conservatism in the heat transfer model.
$^ "
Throttling of the HPSI pump (s) is required to enter shutdown cooling 2.
for breaks small enough that the Safety Injection System can keep the r
The operator will gradually close valves RCS pressurized above 400 psia.
in the HPSI pump discharge lines until the RCS pressure is reduced below Remote position in-400 psia, as indicated by the pressurizer pressure.
5f dicators on these. valves and HPSIP flow meters (F-311, 321, 331, 341) wil' aid the op.'rator during this process.
It is understcod that the conflict-ing infor.. tion contained in T0-CE-660 has been rectified by subsequent i
correspondence.
J i'
3.
The small break LOCA analysis does not account for operation of the Chemi cal and Volume Contro.1 System (CVCS).
The small break LOCA spectrum includes analysis of break sizes with areas 4.
in the range of 0.5 to approximately 0.000S f:2 The icwer limit is chosen consistent with that break size at which the leak flow is matched 3
by the charging flow at full power conditions.
For a given cold leg qu' break, increased leakage assumed from the reactor cociant pump seals define a larger equivalent break size with the transient response "C' "~
2 modated in the above defined spectrum.
Breaks larger than 0.5 ft are accommodated in the large break spectrum.
4 i 1
06s I
1 OC 1
OG I
d
~
V m.
f s
g "h-.
--__mm.
~ ~*****
.*m m,ds Q,, Q [
r I:
l s',
TD-CE-900 2
Transient Analysis _:
tlith natural circulation referring to the initial subccoled portion of the small break LOCA and boiling referring to the c S.
the core and steam generator are:
Steam Generator Heat Transfer RCS Fluid State Core Heat' Transfer Forced Convection
- 1) Subcooled Forced Convection Two Phase Flow With 2)
Saturated Flow or Pool Nucleate Condensation Boiling The time the transiffs from a subcooled forced convection con boiling (saturatedcondition)isachievedinthgcorewillvarywith break to about 40 break size from about fly tsjconds_for a 0.5 ft 2 break.
seconds for a 0.0S ft The hot and cold leg fluid temperatures at the time the transition from a; subcooled to a saturated condition in the core occurs wil 6.
what with break size.to 0.05 f t, the hot and cold leg temperatures will be very 2
2 This information 0.5 ft to the temperatures at initial full power ccnditions.
will be provided at a later date when the Final Safety Analysis has been
!.v completed.
Operation of the main coolant pumps during a small break LOCA ben Core the transient response by delaying or preventing core 7.
The two phase level recedes below the top of the active core region.
pressure drop generated by operation of the main coolant pumps can support a higher fluid level in the inner vessel reg with the coolant pump shut off, so that the potential for core uncovery-occurr.ing during the transient is either minimized or prevented.
a f.
If the break is isolated during the transition to boiling or during the boiling mode, the high pressure safety injection p j
8.
j ;i The unit operator will have at least the folic ing instrumentation avail-These in-l 9.
(1) pressurizer pressure; and (2) RCS te..crature.
l struments are redundant safety channels supplied with emergency power able:
and qualified for post-accident service.
i I
1 044
, 't.
~
^
"' K 7
---.- w _...
^ ~ ' "
~
G-..
e
e
)
=.
a a
~
-,~
T0-CE-900 Recovery _:
The safety injection pumps and the auxiliary feedwater pumps are auto-matica11y actuated by the safety injection actua 10.
the LOCA, the operator initiates cooldown toStea:n is relieved through the supplying feedwater and relieving steam. turbine bypass sy dump system if power is unavailable.
high pressure safety injection (HPSI) pump discharge lines are re so that the total injection flow is divided equally between the hot and the cold legs to maintain core cooling and boric acid flushing.
At eight hours after the LOCA, if RCS pressure has remained above a
specified pressure, the RCS is filled with liquid wa Cocling of the RCS with the steam generators con-The can be established.
tinues until the shutdown cooling entry temperature is achieved.
HPSI pumps are then throttled until RCS pressure is reduced to the Next, all injection ficw is shif ted back down cooling entry pressure.
to th: cold legs and shutdown cooling is initiated.
If RCS pressure at eight hours has fallen (or remained) below the sp pressure, the break may be too large for absolute assurance that pro suction is available for the shutdown cooling mo will both cool the core and flush the reactor vessel indefinitely.
The heat removal mechanism during the transition from boiling (satura RCS condition) to natural circulation ( a subcooled RCS 11a.
given in the response to Question 5.
The time transition from boiling to natural circulation (when t L'
lib.
refills) will vary with break size.
transition will occur from approximately two to ei
{
to 0.05 ft, respectively.
2 Refilling of the RCS~and hence the steam generators is accomplish The HPSI lic.
injection from the high pressure safety injection pumps.
pumps have sufficiently high shutoff head to continue injecting u the RCS becomes refilled and subcooled.
As des-Shutdown cooling is not to be initiated in the boilin 12a.
mode (RCS pressure is less than a specified pressure, eight hour pos LOCA, indicating a large break) hot side / cold side injection will coo
. 2..f the core indefinitely.
pvtdl 7c+.jo i-w a.ac r.
I 01 64 e
N gg
T0-CE-900 (continued):
Recovery CS The unit operator will have at least the pressu s to 12b.
Question 9.
The LPSI pumps will take suction on the hot legs in the nomal mann of shutdown cooling initiation.
cc &,@-72c.
13a.
The high pressure safety injection would con 13b.
The total injection flow-will provide some additional injection flowrate. rate followin i d for the HPSI pumps.
please refer to the response to Question 2.
13c.
The SIT's must be vented to reduce pressure to 400 psig by using 643.
In addition, and/or SI-613, 623, 633, 14.
will be closed w en the valves SI-605, 606, 607, 608, h
the SIT discharge valves, SI-614, 624, 634, 644,RCS As discussed in the response to Question 10
/
15.
If the RCS and there is assurance shutdown cooling can be initiated.
\\
large pressure is less than a specified pressure, the break may be too for absolute assurance that proper suction is av The capacity will both cool and flush the reactor vessel indefinite ling 300,000 gallons.
capability, based upon the interface requirment of I
General _:
The d.etailed emergency procedures are not available at this time.
16.
C-E is currently performing the analysis required to obtain this mation and will respond in separate correspondence.
17.
stion 18 will be answered As agreed upon during the telecon with TVA, ' Assuming two condensd.;
l 18.
2 break.
gallons in the RWT, the time to RAS l
for t!'te 0.1 ft HPSI pump operating, and 469,000 wculd be about 48 to 55 minutes.
1 s ;
i
(
066
- i 3
i t
g
's
- +
4 L ^
^M
= :-
6.
l f
s ::
- 1.
- TD-CE-900
..y General (continued):
The mechanical stresses imposed on the steam generator tubes by the pressure pulses associated with the transition from natural circulation 19.
to boiling and back to natural circulation, are no greater than the stresses imposed by the flow transition during the loss of coolant acci-dent.
As discussed in the response to Question 10, if the RCS pressure is above 20.
a specified pressure eight hours following initiation of the transient, the RCS has refilled (and is in the natural circulation mode) and shut-If the RCS pressure is less than a spe-down cooling can be initiated.
cified pressure (tile RCS is not refilled and is in the boiling mode) eight hours following the initiation of the LOCA, the break is too large.
How-for absolute assurance that shutdown cooling can be initiated.
ever, hot side /. cold side injection alone will cool and flush the reactor Thus, establishing natural circulation (a sub-vessel indefinitely.
cooled RCS fluid state) is not a requirement to recover from the tran~
sient.
v i
i t
- i i
.: )
i
(
1 067 i
_.L_.
~
e6
~ ~" VT 5= n s s
-m
~
, -,,;g
- =
y,
- - - - - - - - - - - - - -