ML20024C608
| ML20024C608 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/31/1976 |
| From: | Cartin L, Hill J, Parks C BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20024C601 | List: |
| References | |
| TASK-07, TASK-7, TASK-GB BAW-10075A, BAW-10075A-R01, BAW-10075A-R1, GPU-2018, NUDOCS 8307120870 | |
| Download: ML20024C608 (28) | |
Text
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BAW-10075A. Rev 1 Topical Repor:
March 1976 e
I L
F MULTDIODE ANALYSIS OF SMALL BREAKS FOR B&W's 177-FUEL-ASSEMBLT NUCLEAR PLANTS WITE RAISED r
Lo0F AIRAm ud.n AND I!rf1'N t VENT VALTES t
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L. L Cartin J. E Rill C. E. Parks 7
l BARCDCE & WILC01 Power Generation Group Nowlear Power Generacion Division P. o. B.: 126o D29378 Lynchburg. Virginia 24305 8abcock & Wilcox 1
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MULTINODE ANALYSIS OF SMALL BREAES FUE El.W's 177-FUEL-ASSEMBLT NUCLEAR PLANTS WITH RAISED I..!
LOOP AREA %1 MENT AND INTERNALS VENT VALVES
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Babcock & Wilcox Power Generation Group kelaar Power Generation Division Lynchburg, Virginia Report BAW-10075A, Rev 1 March 1976 Multinode Analysis of Small Breaks for B&W's 177-Fuel-Assembly Nuclear Plants With Raised Loop Arrangement and Internals Vent Valves - Revision 1 L. 1. Cartin, J. M. Hill, C. E. Parks Eev Words
<inode Analysis. Nuclear Planc, Small Breaks ABSTRACT h icinode analyses were conducted for several small breaks in the reactor coolant system of B&W's 177-fuel-assembit nuclear plants with a raised loop arrangement and internals vent valves. h amitinode blow-down code CRArt was used to evalusta the hydrodynamics and transient water inventorias of the reactor coolant system.,The FOAM code was
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used to compute a swell level history for the core, and THETA 1-B was used to perform transient fuel pin thermal calculations. Curves show-ing parameters of interest are presented. The results of these analy-I ses are acceptable within the guida W== set forth in the Final Acceptance Criteria.
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INTRODUCTION.........................
1-1 2.
St20fARY AND CONCLUSICNS................... 2-1 3,
ANALTTICAL METHOD......................
3-1 3.1.
General........................ 3-1 3.2.
CIAFT Model......................
3-1
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3.3.
FOAM Model....................... 3-3 l
3.4.
THIZAl-B Model.....................
3-4 0
4.
m t m cu. a:SutTs...................... 4-1 4.1.
Explanation of Curves................. 4-1 ii 4.2.
0.3-ft2 Break at Pump Discharge............
4-1 4.3.
0.3-ft2 Break at Pump Suction.............
4-2 4.4.
0.1-ft2 Break at Pump Discharge....... -.... 4-3 i
4.5.
0.04-ft2 Break at Pump Discharge...*... *...... 4-3 5.
BZFERENCES.......................... 3-1 Iw List of Fiaures 5
Figure i
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3-1.
CRAFT Model Noding Diagram.......'......... 3-6
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3-2.
POAM Power Shape..................... 3-7 4-1.
Leak Flow Ys Time for 0.3-ft2 Break at Pump Discharge'..
4-4 L-4-2.
Inner Yassel Liquid Volume Ys Time for 0.3-fc2 Break at Pump Discharge....................... 4-5 4-3.
Normalized Power Ys Time for 0.3-ft2 Break at Pump L
Discharge........................
4-6 i
4-4.
Pressure Ys Time for 0.3-ft2 Break at Psamp Discharge... 4-7 4-5.
Quiet wa:er Level rad Mixture Height in Core Ya Time l
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for 0.3-ft2 Break at Pump Discharge...........
4-8 4-6.
Hot Spot Cladding Te.sperature Ya Time for 0.3-f t2 Break i
at Pump Discharge.................... 4-9 4-7.
Hot Spot Fluid Temperature Ys Time for 0.3-ft2 Break at Pump Discharge.................... 4-10 4-8.
Not Spot Heat Transfer coefficient Ys Time for 0.3-ft2 Break at Pump Discharge.................
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Revision 1 (10/30/75) 1.
DITECDUCTICE This topical report evaluates the effectiveness of the emergency core cooling systems for 3&U's nuclear steam systems with the following features:
1.
177 fuel assemblies.
2.
A loop arrangement in which the steam gen-erators are raised relative to the reactor vessel.
3.
Four 14-inch-diameter internals vent valves.
4.
Nine-inch-diaaster flow restrictors in the core flooding nozzles of the reactor vessel.
5.
Cross-connected I.PI systems.
m Thisanalysisinvestigatesloss-ofboolantaccidentsresulting from small breaks in the reactor coolant system with a core power level of 2772 NWt. The ar.alytical results are also conservative for plants of s4=itar design operating with Mark-C (17 x 17) fuel assemblies or at lower power levels.
A small break is defined as a rupture of the reactor coolant sys-tea with a cross-sectional area less than 0.5 ft. For this analysis, 2
break areas ranging from 0.04 to 0.3 ft2 were considered. Analyses of
" larger" small breaks, such as the double-ended rupture of a core flood-2 ing tank line (0.44 f t ) and a 0.5-ft2 break, are reported in the appli-cant's FSAE and B&W topical report BAW-10105, respectively.. These l1 1
analyses produced acceptable results; thus, the full spectrum of breaka has been considered.
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SincunY AND CONCLUSIONS The results obtained from the spacersna of small breaks analyzed in section 4 show that the cladding will undergo only a moderate in-crease in temperatura during small loss-of-coolant accidents. Because
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the peak cladding temperatures are low, no pocencial for cladding swelling or metal-water reaction exists. Therefore, the core geometry is unchanged and amenable to cooling. Long-term cooling is established as the injection rate matches the leak :ste and the core is covered I
f, with a steaarwater mixture.
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For the breaks analyzed, all conditions of the Final Acceptance
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criteria are anc. Conformance of this analysis to the Final Acceptance 1
Criteria is demonstrated in Appendix A of BAU-10104, Revision 1.8 The results are simmerized as follower a
Break size, Claddina temo. F 2
Time at which long=
ft / location Initial Peak term coolier available, s j
0.3/ pump disch 700 1090 385 0.3/ pump suction 700 824 490 0.1/ pump disch 700 700 1800 1
0.04/ pump disch 700 700 3500 In view of these results, it is concluded that'the present design of the ECCS is more than adequate to control the consequences of loss-of-coolant accidents resulting from==a11 breaks in the reactor coolant system.
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ANALYTICAL METHOD 31 General The These analyses vers performed using three computar codes.
i code is used to calculate systen hydrodynamics during the acc a 1
The core liquid height, pressure, and power as calculated by CRAFT code, which computes a core sixture lev-dent.
S CRAFT are input into the F0AM Outputs from both codes are combined to generate input for the THETA 1-B" code, which is used to calculate the cladding temperature el.
The analyses for each break size are carried out until the h
ore six-transient.
injection rate equals or exceeds the leak rate, so that t e c d
ture level is increasing and long-term cooling is ensure.
3.2.
CRAFT Model The CIAFT code is used to analyse the system hydrodynamics during l
the liquid inven-the blowdown portion of the accident and to calcu ate A 16-node CIAFT cortes in the reactor vessel throughout the transient.
l andel, shown in Figure 3-1, is used to simulate the reactor coo ant It consists of 14 nodes representing the primary systen, one h t.
s: stem.
node for the secondary system, and a single node for the cont =
5&W topical report BAW-10052,5 Multinode Analysis of Small 3 reeks for B&W's 2568-MWt Nuclear Plants, describes a comparison of results break calculated by a small-leak CRAFT model and by the 2
for a 0.5-fc k
more detailed model, which was used in the analysis of large bru s.
Both andels predicted nearly identical results when consistent assusp-Since the nodal representation used for this anal-
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tions were applied.
ysis is more detailed than the model used in BAW-10052, an adequate
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Before a postulated pipe failure, the following assumptions are e
r applied:
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y For all nodes except the core, the distribution factor is taken as 1.0, which implies that the steam is considere[ co be uniformly mixed within Within the core volume detailed FOAM calculations the mixture volume.
have shown that, due to axial heat deposition, a distribution factor of 2.0 is appropriate for the blowdown analysis.
3.3.
70AM Model Me small-leak transient is characterised by a %uest-steady-state yor part of this time, the system for the majority of the transient.
During this period, the upper regions of the core may become uncovered.
lower regions of the core will be in a pool-boiling heat transfer mode with a two-phase fluid for a sink, and the upper regions will be cooled The by the steam being generated by the covered portion of the core.
was developed FOAM code, as documented in B&W topical report BAW-10064,3 to calculate two-phase mixture heights during this time period.
From the CIAFT code, core liquid height, core power, and core pressure as functions of time are input into FOAM to calculate the The core liquid height is calculated from the inner vas-swell level.
sel liquid volume, which is the sua of the liquid volume in the lower For the CRAFT plenua, core, core bypass, upper plenum, and upper head.
model used, this volume is the sum of the liquid volume in nodes 13 and The volume calculated is distributed in the inner vessel column so 2.
The r=== N ec of the that the lower pianum will be filled with liquid.
water is in the core and is used to calculate the quiet core liquid j
This approach is conservative since no credit is taken for the level.
aiature that will exist in the lower plenum due to primary metal heat
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This leads to less water in the core than is addition and flashing.
expected and thereby results in lower swell levels.
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and pressure are used as input to THET.A during all phases of the acci-danc.
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ANALYTICAL R11ULTS 4.
of the breaks consid-i This section presents a detailed evaluat cehenomena involved.
Size effects ered, along with explanations of the pand one esse was analyzed at the h
e were analyzed at the pump disc arg,ffects.
pump suction to determine location e Ernianation of Curves _
d to aid in understanding 4.1 The following explanations are provide he curvess the parameters illustrated in t ge core mass flux core mess flus - This is a ploc of the smoothed avera20% for inpu as predicted by CRAFT and reduced by alized thermal power as calcua Core neve_r - This curve indicates the norm d by CIAFT.
laced by CIAFT.
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Inner vessel licuid volume -
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ddfng temperature as calculate d
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temperature calculated for this acc Break at Pumo Discharte_
2 k is 4.2.
0.3-ft break, considered a medium-sized small brea,of 2
The 0.3-ft i
During the characterized by a ra,11d depressurizat oni g within one second.
with the reactor sad pump trips occurr nwater volume is substantial, but next 200 seconds, the reduction in coret 185 seconds, the liquid level k
after activation of the core flood tan s a is observed.
At 282 seconds, in the core increases; a continual increaseThe pressure and inner vessel liq
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show that cladding temperature the as: break.
i ns anks begin flowing atwell level calculat oentire transient.withis the sixture is by No re-Transient saircure for the transfer will thus eith two-phase cessary because heatcladding temperature Lons-rature.
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REFERENCES I
W. L. Bloomfield, et al., 3CCS Evaluation of B&W's 177-FA Raised-Loop NSS, BAW-10105, Babcock & Wilcox, Lynchburg, Va., June 1
1975.
CRAFT - Description of Model for Equilibrium LOCA Analysis Program, BAW10030, Babcock & Wilcox, Lynchburg Ya., October 1971.
3 B. M. Dunn, C. D. Morgan, and L.1. Cartin, Multinode Analysis of Core Flooding Line Break for B&W's 2568-We Internals Vent Valve Plants, BAW-10064, Babcock & Wilecx, Lynchburg, Va., May 1973.
C. J. Bocavar and T. W. Winaisser THEA1-B - A Computer Code for Nuclear Beactor Core Thermal Analysis, IN-1445, Ilabo Nuclear Corp.,
February 1971.
5 C, E. Parks, B. M. Dunn, and 1. C. Jones, Multinode Analysis of Small Breaks for B&W's 2568-MWe Nuclear Plants, BAW-10052, Babcock & Wilcox.
Lynchburg Ya., September 1972.-
6 J. F. Wilson, 1. J. Granda, and J. F. Peterson, "The Velocity of Bising Steam in a Bubbling Two-Phase Mixture," ANS Transactions, 5, (1962) p 151.
7 C. D. Morgan and B. S. Kao, TAIT - Fuel Pin Temperature and Gas Pres-sure Analysis, BAW10044, Babcock & Wilcox, Lynchburg, Va., April 1972.
s B. M. Dunn, se al., B&W's ECCS Evaluation Model, BAW-10104. Bav. 1, Babocck & Wilcox Lynchburg, Va., October 1975.
B29428 5-1 Babcock & Wilcox
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