ML20002C410
| ML20002C410 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/28/1976 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20002C404 | List: |
| References | |
| NUDOCS 8101100267 | |
| Download: ML20002C410 (6) | |
Text
i r)
.\\
-~
O BIG ROCK POINT PLXIT l'
ANALYSIS OF RING SPR.\\Y LINE BREAK WITH FEEDWATER ADDITION l
General Electric Company April 28, 1976 Introduction Ccusumers Power Company has asked General Electric Company to respond to an NRC request which states in part as follows:
" Provide an analysis of the ECCS performance which properly demonstrates that in the event of a break in the core ring spray line, the feed-water system and the flow through the core spray nozzle will reliably provide sufficient core cooling water unless adequate spray distribution of the nozzle has been demonstrated."
.This analysis shows that sufficient core cooling water is provided under the conditions described below.
Description of the Analvsis A complete severance of the ring core spray line just outside of the reactor pressure vessel was assumed.
I==ediately after this hypothetical event, coolant from the regfan above the core will flow backwards through the thirty-six(l)* spray nozzles, around the ring-shaped sparger, down the riser pipe which connects the vessel nozzle with the ring sparger, and out the
~
vessel nozzle through the broken pipe into containment.
i diameter (2). ype 1/2 G1590 nozzles has a flow orifice which is 7/32" in Each of the t This orifice linits backward flow ti: rough each nozzle in the event of the ring spray line break. The total orifice area of all 36 nozzles 2
is 0.009392 ft', or 1.35 in.
It is assumed that critical flow is estab-lished at tbase orifices af ter the break, and that the approved Moody Critical Flow model in the SAFE code determines the break flow rate.
Note that earlier analyses of the ring spray break, in the absence of data on the 36 spray nozzles, had used a critical flow area of 0.03903 ft2 or 2
5.63 in, which corresponded to the spray line flow cross-sectional area at the vessel nozzle.
Use of the correct flow-limiting area in this analysis predicts a slower depressurization transient than was predicted in earlier analyses.
- Numbers in parentiieses refer to references at the end of this report.
~
N t
ppjJcQ96Y
/
l i
2 The elevr'. ion of the break was similarly corrected to the level of the lowest 7
spray nozzle instead of the level of the spray line vessel nozzle as used in earlier analyses. The new elevation is 4.33 ft. above the top of the core; the old elevation was 3.77 ft.
Feedwater Addition In modelling the behavior of the feedwater pumps during the. ring spray line LOCA, conditions which could cause feed pump trips were considered. These conditions include:
a trip on low suction pressure, which could result from a low hotwell Icvel or a condensate pump trip; or a trip on high current, which could result from runout flow.
Since any of these conditions could cause a feed-pump trip following a ring spray break, and since the exact time of such a trip is not known, a bounding assumption was made:
the feed-pumps trip at the time of-the break.
Following' trip, the feed-water flow coasts down to zero in one second, as was assumed in earlier analyses.
Re-establishment of feed flow would require operator action.
It was further assumed that the operator restarted one of the two f eed-purg 10 minutes af ter the LOCA, and that full flow f rom the operating feed-pu-p is attained 10 seconds after restart.
Thereafter the operator regulates feed flow at t5 normal (pre-LOCA) value until the normal steam drum water level is cec-in:d.
{
The temperature of the feed-water delivered to the reactor vessel following(3),
feed-pump restart is taken as 1000F, in accordance with a previous analysis Single Failure Consideration Pric-to making computer runs, it was decided that the worst single failure would be ene of the four Reactor Depressurization System (RDS) valves failing to open, in accordance with the previous analysis of the ring spray line break (4).
However.
the results of the analysis indicate that the RDS does not actuate; henc.. rostulating a single failure in this system does not ems, which are the nozzle core spray and the feedwater affect the acLive s.
systems. A sing 1_ :. lure could be postulated in either of these systems.
Such a single failure would completely disable either system, not merely degrade the system.
Since the basis of this analysis as stated in the NRC request (see page 1) implies that credit may be taken for flew through both the nozzle spray and feedwater systems, no further single failure was postu-lated in this analysis.
Results
(
Following a ring spray line break, the liquid level in tSe primary system is predicted by the SAFE code as shown in Figure 1.
Theminimda1cvelis Y
FHure 1 i
__ ---__ B I G ROC 4 3T i
i I
RING SPRRY BREAK 0.00939 FT2 60" i
VESSEL WATER LEVEL 1
t10.
7
-L w
LLJ TA F m
_J
_88F LLJ y
gg*
o47.376 1Z0719' g
"1l3 snew 8 (CS45
~
y C3 t,23 cso3c n:
C2 12 J 2
,,:x:3 A
g-I:b O
^
.Jt3 O.
IL---
- - - - =
O.
25
- 0. 'l 0.8 1.2
- 1. 6x10' 1IME
{ SECONDS) r-
i 3
(-
31.98 feet at 600 seconds, just before the feed-pump is restarted.
Since the top of the core is at 28.78 feet, there is at least 3.20 feet of water above the core during this LOCA.
Because the core is kept covered with liquid, there is no heatup phase to this LOCA, and cladding temperatures remain close to their operating values.
It is concluded that sufficient core cooling is provided by the feedwater system in the event of a break in the core ring spray line.
Y i
f e
G e
i
o..
(
t 4
(.
References 1.
General Electric drawing of the ring spray sparger, No. 212E456.
2.
GE internal memo by KH Sun, " Summary of Existing Information for GE Domestic BWR/l Core Spray System", March 31, 1976.
3.
Submittal from Consumets Power Company to the USAEC dated June 28, 1974; Figure 1.
4.
Submitta". from Consuoers Power Company to the USNRC dated July 25, 1976; Appen#.1x A, Table 4.
O t h r
1 1
4 "
- ,y
(~
~
t I
a Y
/
r!HCeonu WS U.S. NUCLE AH HEGULATOHY
.1 MIS 310N DOCKCT NUM00 H 2m 50-155 I
NRC DISTRIBUTION roit PART 50 DOCKET MATERI AL TO:
NRC FitOM:
Con., - es, Power Co.
oATE or ocCuMENT Jackson,' Reih. 49201 5-10-76 Ralph B. Sewell oATE RECElVED 5-12-76 klLL TTE n
,0 NOTO RIZ E D PHOP INPUT rORM NUMOEn oF COPWS RECEtVED n IN A t.
OUNC L ASSIFIE D 3 signed 37CC DESCRIPTION Ltr. trans the follow ng:
ENCLOSU RE Request for Change to the Te:h Specs notarized 5-10-76 consisting of changes in the surveillance requirements for core spray systems, changes in the conteinment spray systems & other changes.....
(3 Orig & 37 CC rec'd)'
$3
.Lv:kL0"$r$"w}1Q Trg PLANT NAME:
Big Rock Pt. Elant i
SAFETY FOH ACTION /INFORMATION ENVIRO DHL 5-12-76 i
ASSIGNED AD :
, _l ASSIGNED AD :
{g/
BRANCH CliIEF :
JL(' 4
- e m 49 AJ BRANCH CHIEF :
@ e.g PROJECT !!ANAGER :
PROJECT MANAGER:
y LIC. ASST. :
D;qgg LIC. ASST. ;
e INTERN AL DISL RIBUTION VT REG FILE
JLSIFMS_SEEIL_
PLMiT SYSTE"S Fu m n TFru
~
W NRC PDR HElHEMAN TEnESCO ERNST LV I _ & E-[1)
SCHROEDER BENAROYA
_J3ALLARD 4M
_DELD IAINAS SPANGLER i
V GOSSICK & STAPF ENQJNERItJG IPPOLITO
.. __ _.EIP_C MACCAP.Y SITE _ TECH
___ CASE KNICHT OPEPATING REACTORS CAF?!ILL
.L.
JJAUAUER SIINEIL STELLO SYO~ P P
. __ EARLESS PAULICKI HUIJ!AN t
OPERATING TECH PROJ2CT MANAGEMENT REACTOR SAFETY if EISENHlrr SITE ANALYSIS BOYD' ROSS
/
Sl!AO BUNCil VOLIJtER P. COLLINS NOVAK n/
BAER HOUSTON
}0SZTOCZY V SCHWENCER V
J. COLLINS I
PETERSON CITECK p GRIEE'S KREGER MELTZ L.
IlELTEMES AT_& I SITE SAFETY & ENVIIRf SKOVit0LT SALTZMAN ANALYSIS Pg g
gn liUTFERG
~55UTUJ & MUT.LER M
$ lM ll J l 2 l gQ a
samaa Eu 1E EXTEHNAL DIS 1 HIBUTlON 5
J ' J I$onMt Mt g4 ft)) j l L
.I_ LPDR:,Gliarlevoix,EL.
NATL LAB BROOKHAVEN NATL LAB F
TIC REC. V-IE ULRIKS ON (O'"iL) f_Njic LA PDR 4772 ASLB CONSUI.TANTS
.V. -ACES MlSl%G YD S' S' _ __ _.. _
_L__
_