ML20039F408
| ML20039F408 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 12/23/1977 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 BAW-10104, NUDOCS 8201120428 | |
| Download: ML20039F408 (17) | |
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SMALL BREAK ECCS EVALUATION OF B&W'S 205 FA NSS USING THE DECEMBER 1976 ECCS EVALUATION MODEL (BAW-10104, REV. 3) 1.
INTRODUCTION The small break LOCA analysis for B&W's Category 3 plants (205 FA NSS) was reported in BAW-10074A.(
An approval letter on that topical was received from NRC on January 8, 1976.( }
Since that date, the ECCS small break evaluation model has evolved to the present model given in BAW-10104, Rev. 3.( }
As a result, the NRC Staff, in a letter from S.A. Varga to J.H. Taylor of May 10,1977, requested a spectrum of three large breaks and the worst case small break analysis wholly in conformance with 10CFR50.46 and Appendix K to confirm that the break spectrum shape had not changed and to confirm the margins available. The large break portion of this request was submitted to the Staff on Septemper 30, 1977.b0} This analysis herein is the worst case small break portion of the above request. This report addresses the total impact of the above model change on the B&W Category 3 NSS small break evaluation.
The core flood line (CFT) break was shown to be the most limiting small break in BAW-10074A, Rev 1( } because it is one of the largest small breaks and it has the minimum ECC system available to mitigate the LOCA.
B20112042B 010403 r
Presented herein are the results of "a core flood tank (CFT) line break analysis for the B&W Category 3 plants.
B&W's ECCS evaluation model, as defined in BAW-10104(3) was used for the analysis. This analysis demonstrates the conservative nature of the cladding temperature calculations presented i
for the same break in BAW-10074A, Rev. 1.
2.
Method of Analysis The analysis uses the CRAFT 2( } code to predict the hydrodynamic behavior of the reactor coolant system. The noding description of the reactor coolant system used in CRAFT for the CFT line analysis is basically the same as that for the large breaks but with fewer nodes. It is demon-strated in Topical Report BAW-10052(0} that, with the same assumptions, 2
the noding model applied to the small break (0.44 f t CFT line break) produces the same system behavior as the more detailed large break noding model.
The CRAFT model uses 19 nodes to simulate the reactor coolant system, two nodes for the secondary system, and one node for the reactor building.
A schematic diagram of the model is shown in Figure 1 along :rith the node descriptions.
Control volumes (nodes) in and around the vessel are all connected by a pair of flow paths to allow the occurrence of counter-current flow. The break is in the core flood line joining the CF nozzle. The Wilson, Grenda and Patterson average bubble rise model is used for all nodes.
Within the core region, however, a multiplier of 2.38 is applied to the calculated bubble rise velocity. Appendix F of BAW-10104 demonstrates that a multiplier of 2.38 in CRAFT 2 gives a mixture height within + 2% of that predi2ted by FOAM.
Thus, no 70AM analysis will be needed if the CRAFT 2 mixture level remains above the core by 2% of the active length.
If the t
core does uncover to within 0.25 f t above the active core, the cladding heatup would be calculated by the procedure outlined in Section 5.2.3 of BAW-10104.(3)
The following assumptions are made for conditions and system responses during the accident:
1.
The reactor is operating at 102% of the steady-state power level of 3800 MWt.
2.
The leak occurs instantaneously, and a discharge coefficient of 1.0 is used for the entire analysis.
Bernoulli's equation was used for the subcooled portion of the transient, while Moody's correlation was used in the two-phase portion.
3.
No offsite power is available.
4.
The reactor trips on low pressure at 1965 psia.
5.
The safety rods begin entering the core af ter a 0.65 second delay from the time the reactor trip signal is reached.
6.
The RC pumps trip and coast down coincident with reactor trip.
7.
One complete train of the emergency. safeguards system fails to operate, leaving one CFT, one HPI system and part of one LPI system to be available to provide ECC fluid to the vessel.
8.
The auxiliary feedwater (FW) system is assumed to be available 40 seconds after loss of reactor power.
9.
ESFAS signal error band is considered in the analysis to signal the actuation of the HPI system.
- 10. The linear heat generation rate in the hot pin is the maximum allowed by the Technical Specifications.
1 3.
Results of the CFT Line Break Transient This section presents a detailed evaluation of the CFT line break considered and explanations of the phenomena involved.
3.1 Explanation of Curves The following categorical explanations are provided to aid in under-standing the parameters illustrated in the curves:
Core Power:
This curve indicates the normalized thermal power as calculated by CRAFT 2.
Core Flow:
This curve represents the total flow rates of core paths 1 and 2 of Figure 1.
The curve shows flow rates mainly during the flow-con-trolled part of the transient.
Pressure:
This is the pressure at the top of the core node as calculated by CRAFT.
In this analysis, the core node includes the core, upper plenum, upper head, and core bypass.
Boil-Off Due to Decay Heat:
The liquid boil-off rate is given in terms of the equivalent amount of HPI or HPI + LPI ECC injection rate needed to dissipate the core decay heat. Mathematically, Boil-off rate = (core decay heat rate)/(h - h,)
g where h = enthalpy of saturated steam at core pressure, g
h,= enthalpy of injected water.
g Inner Vessel Mixture Height:
This curve shows the mixture height in the core node as calculated by the CRAFT code. The lines spanning the curve indicate the top of the active core, the hot les regions, and the vent valve region.
Core Liquid Level: This curve, in contrast to that for the inner vessel mixture height, shows the effective core liquid height and volume with the lower plenum filled with a mixture at the void fraction calculated by CRAFT 2.
This volume is representative of the liquid volume within the core node that would be used to calculate the mixture height within the core.
3.2 0.44 Ft CFT i.ine Break
(
The break is assumed to oe at the CFT line nozzle joining the reactor 2
vessel and is limited in area to 0.44 ft by the nozzle insert in the CFT line. Node 13 in Figure 1 is the break node, and the analysis takes credit for one CFT, one HPI pump, and a portion of one LPI pump available for core cooling. The LPI flow is available because it is cross connected l
to each flooding line*with flow limiting devices. To ensure a conservative calculation, no credit for LPI was taken until the system pressure fell below 165 psig. Table 1 preseaIts the sequence of events for the CFT line break.
Figures 2 and 3 show core power and core flow rate. Rapid initial depressurization (see Figure 4) causes reactor trip and the start of the
.RC pump coastdown within the first second.' Flashing of system liquid slows t
the depressurization while the steam generator continues to remove energy from the primary coolant, thereby helping to decrease the pressure. The lower pressure limit of the ESFAS setpoint error band is reached by about i
15 seconds, which initiates main feedwater and steam isolation procedures and signals the actuation of HPI. At about 40 seconds, the RCS pressure drops below the secondary steam generator pressure, and heat removal to the sccondary side drops off sharply and becomes a source of heat to the
~
primary, causing a slower depressurization.
System flow has degraded so e
w
.-,y w
g.
yv.
-6 that core flow is predominantly due to na.tural circulation, and the quiescent period of the tradsient begins.
The HPI system provides makeup starting at about 50 seconds, aiding depressurization.
CFT flow begins at 135 seconds, aiding further depressurization, but the diminishing leak flow (Figure 5) slows the depressurization rate. Figure 6 is a plot of ECC injection rate. As shown in Figure 7, the CFT flow into the reactor vessel turns around the decreasing inner vessel mixture height and core liquid level.
At 380 seconds, the system pressure falls below 165 psig and LPI flow begins to enter the reactor vessel. Long term cooling is ensured in that by 385 seconds the high-pressure injection rate plus the low-pressure injection rate exceeds the boil-off due to core decay heat. Figure 7 is a plot of inner vessel fluid inventory.
It shows that, while much of the core liquid inventory,is depleted, the mixture level predicted by CRAFT remains at a level where it is able to spill into the hot legs and, for some of the time, through the vent valves.
Thus, no cladding temperature transient will occur since the core is always covered with a mixture and the RPI rate plus the LPI rate has exceeded the boil-off ensuring long-term cooling capability.
4.
Summary and Conclusions The CFT line break analyzed in this report shows that the core remains covered throughout the transient. During the initial period when the transient is flow-controlled, sufficient flow is maintained so that CHF does not occur, and nucleate boiling heat transfer predominates.
Since the core remains covered by a mixture, pool film boiling will be maintained during the quiescent period of the accident. This heat transfer
^
mechanism is sufficient to maintain the cladding temperature within a few degrees of the fluid saturation temperature. Therefore,for the maximum linear heat rate covered by the Technical Specifications, the transient cladding surface temperature will never exceed its initial value of approximately 660F, no metal-water reaction will occur, and the core geometry will remain coolable since no cladding rupture will occur. Long-term cooling is estab-lished as the HPI and LPI pumped injection systems provide fluid in excess of the boil-off rate due to core decay heat. Thus, the five Acceptance Criteria in 10CFR50.46 are met.
The new modeling techniques used in the CRAFT 2 analyses for the present studies show improvement in the core performance when compared with the results of the same break reported in BAW-10074A, Reference 1.
Therefore, if all the small breaks reported in this report were re-analyzed with the present model, the same trend of improvement in core performance would be realized. Thus, the present analysis, in conjunction with the analyses
,of BAW-10074A, Reference 1, provide a suitable small break spectrum for demonstration of compliance of the ECC system with the five Acceptance Criteria in 10CFR50.46.
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_5. References [.
l..
R.C. Jones, et.al.,
Multinode Analysis of Small Breaks for B&W's 205 Fuel Assembly Nuclear Plants with Internal Vent Valves, BAW-10074A, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, March 1976.
2.
Letter from A. Schwencer (NRC) to K.E. Suhrke (B&W), January 8,1976, (Approval of BAW-10074).
3.
B.M. Dunn, et.al., B&W's ECCS Evaluation Model, BAW-10104, Rev. 3, August 1977.
4.
Letter from J.H. Taylor to S.A. Varga of September 30, 1977.
5.
R.A. Hedrick, J.J. Cudlin and R.C. Foltz, CRAFT 2 - Fortran Program for Digital Simulation of a Multinode Reactor P 3 ant During Loss of Coolant, BAW-10092, Rev. 2, Babcock & Wilcox, April 1975.
[
6.
C.E. Parks, B.M. Dunn, and R.C. Jcnes, "Multinode Analysis of Small Breaks for B&W's 2568 MWt Nuclear Plants, BAW-10052, Rev. 1, Babcock & Wilcox, Lynchburg, Va.,
October 1975.
7.
J.F. Wilson, R.J. Grenda, and J.F. Patterson, "The Velocity of Rising Steam in A Bubbling Two-Phase Mixture, "ANS Transactions, 5 l
(1962).
8.
B.M. Dunn, C.D. Morgan, and L.R. Cartin, Multinode. Ana. lysis of Core.
Floodieg Line Break for B&W's 256 8 MWt Internals Vent Valve Plants, BAW-10064, Rev.1, Babcock. & W1.lcox, October 19 75 (tha FOAK code is 1
1 discussed in this reference).
S
s 6.
Computer Data Version Name Version Date Run Name Run Date CRAFr2, Version 8.2 9/19/77 CF502YM 12/20/77 7.
List of Figures 1.
CRAFT 2 Noding Diagra m for Core Flood Line Break 2.
Core Power for CFT Line Break l
3.
Core Flow for CFT Line Break 4.
Core Pressure History, CFT Line Break 5.
Leak Flow vs. Time, CFT Line Break 6.
ECC Injection Rate, CFT Line Break 7.
Inner Vessel Fluid Inventory, CFT Idne Break i
I 4-D 4
4 e
e 4
l 4
TABLE 1 1
Sequence of Everts for CFT Line Break r
Break 0.0 RPS Iow RCS Pressure Trip
- 0. 3 Safety Rods Begin to Enter Core 1.0 ESFAS Actuation on Low RCS Pressure 15.0 HPI's On 50.0 Intact CFT On 135.0 LPI On 380.0 Long-Term Cooling Ensured (ECC Injectioti Rate Exceeds Boil-Off Due to Core Decay Heat) 385.0 T
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l Figure 1.
CRtET2 NODING DIAGRAM FOR CTT LINE BRDK 1@1 1h1 21 15 IT 4
2 b
Y.
IB 6
>)e
_D - @g EfS i i.
I (1,2) 28 II II 2
g Y
l T
33
~
g NPI 38 u
NOTE: ADDITIONAL DATA NOT $HCNN 12 DN DIACRLE ARE:
33 T
23
~
h
@ G
- D NCOE 22 15 CCNTAINutNT NODE G
LPl 2
g is LEAL PATM FRou PATM BREAK TO CCNTAINNENT h 13 RETURN LEAK PATM FR 3 PATM CONTAINNENT TO 8REAK N3DE h REPRESENTS CONTAINEENT PATM SPRAT SYSTEE N00!_N0 IDENTIFICATION PATM NO.
IDENTIFICATION 2
I 00tNC0uER I.2 CORE 2
LCIER PLENUE 3,4,18,19 NOT LES Pirtut 3
CORE, CCRE ITPAS$ UPPER 5,20 HOT LEG, UPPEg PLENUN, UPPER HEAD 6,21 SE TU8E5 4,14 MOT LIG PIPING T,22
$G LCIER HEAD 5.15 STIAE CENERATOR UPPER I
CORE BTPAss HE AD, SG TUBE 3 (UPPER HALF) 9,13,24 COLD LIG PlPING E,18 3G TUBE 5 (loser HALF) 10,14,25
- Puun 8,18 3G LCeER HEAD 11.12,15,18,26,2T COLD LEC PIPING I,11.19 COLD LES PIPING (PUur SUCil0N) 1T.31 DOWNCOuER 10,12.20 COLD LEC PIPING (PUMP Dl3 CHARGE) 23 LPI 13 UPPER CCfNCC4ER 20,29 UPPER 00fNC0eER (ABOVE THE ( OF N0ZZLE BELT) 30 PRES $URil!R I
21 PRES $UtilER 32 ViNT VALVE
- 2 CCNTAlksENT 33,34 LEAK & RETURN PATH 35,38.38 NPI 3,
CONTAIN4ENT $ PRATS 9
4
FIGURE 2.
CORE POWER FOR CFT LINE BREAK
- 1. 0 0.8 U
5 0.6 a.
E O
W 5
0.4 e
O.2 0.0 i
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i 0
50 100 150 200 250 300 350 400 Time, s.
FIGURE 3 CORE FLOW FOR CFT LINE BREAK 40 30 "o
H x
20 -
e D
H g
10 r
LL 0
-10 I
a I
e t
0 100 200 300 400 Time, s t
---,-n..
FIGURE 4 CORE PRESSURE HISTORY, CFT LINE BREAK 2500 2000 T
Initial System Pressure = 2247 psia
%g 1500 al e
E
[
1000 500 -
1
.a.
O i
e i
i i
e i
i 0
50 100 150 200 250 300 350 400 Time, s
0 0
4 05 3
0 03 KAERB 0
E 25 N
I L
TF C
s 0
E 0
M 2 e I
m T
iT S
V W
O 0
L 5
F 1
KAEL 5
0 0
E 1
R UG IF 0
5 O
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 2
0 8
6 4
2 1
1 m10 3e 8a"-
FIGURE 6 ECC INJECTION RATE, CFT LINE BREAK 800 600 s
ED H
4 In]ection Rate
/
- c:
400
, g<
H I
u.
y l "V l
i:
i 200 Decay Heat Boiloff Rate
%. N..
,, ' ~ ~
_, j{
Injection Rate
...--y-->
L HPI Injection Rate 0
I I
I I
I O
100 200 300 400 500 Time, s
~
FIGURE 7 INNER VESSEL FLUID INVENTORY, CFT LINE BREAK 32 28 f
d Inner Vessel Mixture Height j.
ao 24 s
'd Vent Valve Region E
~ ~ / sm' / / / /
- W ff/ ///////// M ///////// /////
}Q i
/
Y_
/
h
-f/
Hot Leg Region l h
/////////
cc
/ /
/
l 16 h
/////////
/b
/
u.
U 1200 g
Top of Active Core o
I 12
-W Core Liquid Level 800 400 4
l 0
1 I
I I
I O
100 200 300 400 500 Time, s
-