ML20024C756
| ML20024C756 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/30/1975 |
| From: | Dunn B, Rosalyn Jones, Parks C BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20024C754 | List: |
| References | |
| TASK-06, TASK-6, TASK-GB BAW-10074, BAW-10074A-R01, BAW-10074A-R1, GPU-0473, GPU-473, NUDOCS 8307130230 | |
| Download: ML20024C756 (29) | |
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P MULTINODE ANALYSIS OF SMALL BREAES 701 B&W's 205-FUEL-ASSEMBLY NUCLEAR PLANTS WITH INTZBNALS *ENT VALVES
- Revision 1 -
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by R. C. Jones
- 3. M. Dunn C. E. Parks
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BABCOCE & WILCDI Power Generation Group Nuclear Power Generation Division P. O. Ecx 1260 i,,
Lynchburg, Virginia 24505 l
8307130230 760108 l
PDR ADOCK 05000289 i
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Revision 1
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(10/30/75)
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Babcock & Wilcon Power Generation Group Nuclear Tower Generation Division Lynchburg, Virginia Topical Report BAW-10074A, Rev 1 March 1976 Multinoda Analysia of Small Breaks
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For B&W's 205-Fuel-Assembly Nuclear Plants With Internals Vent Valves - Revision 1 R. C. Jones, 3. M. Dunn, C. E. Parks Key Words: Multinode Analysis. Nuclear Plant. Snmil Break 4
ABSTRACT b.
Multinode analyses were conducted for several,small breaks in the reactor coolant system of B&W's 205-fuel-assembly nuclear plants with in-ternals vent valves. The multinode blowdown code CRAFT was used to evalu-ate the hydrodynamics and transient water inventories of the reactor coolant system.. The FOAM code was used to compute a swell level history for the core, and the THETA 1-B code was used to perform transient fuel pin thermal calculations. Curves showing the parameters of interese are presented. These results are well within the Final Acceptance Criteria.
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CONTENTS Page
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1.
INTRODUCTION.........................
1-1 2.
SUMMARY
AND CONCLUSIONS...................
2-1
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3.
METHOD OF ANALYSIS......................
3-1 F
3.1.
CRAFT Model......................
3-1 6
3.2.
FOAM Model.......................
3-3 3.3.
THETA 1-B Model.....................
3-4 P
4.
RESULTS OF ANALYSIS.....................
4-1
,s.
4.1.
Explanation of Curves.................
4-1
/
4.2.
Core Flooding Line Break................
4-1 4.3.
0.5-ft2 Break at Pump Discharge............
4-2 4.4.
0.3-ft2
~
Break at Pump Discharge............
4-3 4.5.
0.1-ft2 Break at Pump Discharge.....
4-4 4.6.
0.1-ft2 Break at Pump Suction.............
4-4 4.7.
0.05-ft2 Break at Pump Discharge............
4-5 I
5.
REFERENCES..........................
5-1 L:
List of Finures Figure 3-1.
CRAFT Model Noding Diagram................
3-5 3-2.
FOAM Power Shape..
3-6 3-3.
Linear Heat Rate Distribution for 12-Axial-Node, THETA 1-B Model Ho t Pin..................
3-7 4-1.
Core Pressure for Core Floeding Line Break........
4-6 4-2.
Normalized Power for Core Flooding Line Break......
4-7 4-3.
Leak Flow for Core Flooding Line Break..........
4-8 4-4.
Inner vessel Liquid Volume for Core Flooding Line Break.
4-9 e
4-5.
Core Mixture Height for Core Flooding Line Break.....
4-10 4-6.
Hot Spot Cladding Temperature for Core Flooding Line Break..........................
4-11 i
4-7.
Hot Spot Heat Transfer Coefficient for Core Flooding Lias Break........................
4-12 4-8.
Hot Spot Fluid Temperature for Core Flooding Line Break.
4-13 Babcock & Wilcox
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1.
INTRODUCTION Ihis topical report evaluates the effectiveness of tha emergency u
core cooling systems for B&W's 205-fuel-assembly (TA) reactor designs with the following features:
1.
Eight 14-inch-diameter internals vent valves.
2.
Cross-connected LPI systems.
3.
HP1 system manifolded into each cold leg.
1" i
t The plant is assumed to be operating at 3760 MWe, and the analysis is conservative for plants of similar design operating at a lower rated power. The analysis is also conservative for plants of similar design with Mark-C design fuel assemblies.
The report presents the results of an analysis of loss-of-coolant accidents resulting from small breaks in the reactor coolant system.
(Small breaks are defined as those breaks in the reactor cociant system V
with less than 0.5-fc2 leak aru.) The CRAFT,1 70AM,2 and THETA 1-B3 go,_
puter codes were used for the analysis.
A spectrum of ruptures is considered: 0.5, 0.3, 0.1, and 0.05-fc2 leak areas. Also, the report includes the double-ended rupture of the l
core flooding line (0.44-ft2 leak area). These ruptures provide an ap-i propriate spectrum for evaluation of the effects of small leaks.
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Revision 1 (10/30/75) i 2.
SUMMARI AND CONCLUSIONS The spectrum of==all breaks analyzed in asction 4 shova that the cladding will undergo only a moderate increase in temperature during small loss-of-coolant accidents. Because the peak temperatures are low, no potential for cladding swelling or metal-water reaction exists. There-7 fore, the core geometry is unchanged and will remain amenable to cooling.
,, Long-term cooling is established once the injection rate matches the p
leak race and the core is covered with a steam / water mixture.
JJ The double-ended rupture of a core flooding tank lina yields a maximum cladding temperature of 1636F. Core geomet y will be maintained in a coolable configuration. Long-term cooling is established by using the same reasoning as in the paragraph above.
I For all breaks analyzed, the conditions of the Final Acceptance Criteria are met.
Conformance of this analysis to the Final Acceptance Crituia is demonstrated in Appendix A of BAW-10104, Revision 1.8 There-1 t
fore, the design of the emergency core cooling system is adequate to con-trol small loss-of-coolant accidents.
Results for the breaks considered are summarized as follows:
Initial Peak Time at shich long-term Break temp. F tene. F cooling is established, s 0.5 ft2 at pump disch 710 1178 200 l
0.3 ft2 at pump disch 710 710 400 0.1 ft2 at pump disch 710 710 2200 i
O.1 ft2 at pump suct 710 710 2550 0.05 ft2 at pump disch
'710 710 2800 0.44 ft2 C7 line break 710 1636 400 2-1 Babcock & Wilcox
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.P 3.
METHOD OF ANAI,YSIS
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These analyses are performed with three computer codes. The CRAFrl code is used to calculate system hydrodynamics during the accident. Ths core liquid height, pressure, and power as calculated by CRAFT are input y
into the FOAM 2 code, which computes a core mixture lavel. Outputs from both codes are combined to generate input fot the THETA 1-B3 code, which 4
is used to calculate the cladding temperature transient.
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i 3.1.
CRA T Model r
The CRAFT code is used to analyze the hydrodynamics during the blowdown portion of the accident and to calculate the liquid inventories in the reactor vessel throughout the transient. The CRAFT model uses 14 nodes to describe the primary system, one node to model the secondary system, and one node to simulate the containment, The noding arrange-ment is shown in Figure 3-1.
In the topical report BAW-10d52,4 it is r
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shown that smaller models than those used for large loss-of-coolant ac-cidents are sufficient to describe the reactor coolant system's hydro-L dynamics. Since the model used for this analysis is more detailed than the one in BAW-10052, the nodal representation used will give numerical
- solution convergence. The following assumptions were made for the anal-ysis:
L 1.
The reactor is operating at a steady-state power level of 3760 MWt before rupture.
l 2.
The break occurs instantaneously. The leak flow is cal-culated by the Moody leak correlation with a C f 1.0.
D 1
I-3.
No offsite power is'available.
I 4.
The reactor trips on low reactor coolant system pressure at 2060 psig.
5.
Rod insertion starts 0.5 second after the trip signal is generated.
3_t Babcock & Wilcox i
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u which implies,that the steam is considered to be uniformly mixed within the mixture volume. Within the core volume, detailed F0AM calculations have shown that due to axial heat deposition, a distribution factor of 2.0 is appropriate for the blowdown analysis.
3.2.
FOAM Model The saan-leek transient is characterized by a quasi-steady-state system for the majority of the transient. Fart of this time, the upper regions of the core may become uncovered. During this period, the lower regions of the core win be in a pool-boiling heat transfer mode with a two phase fluid for a sink, ad the upper regions will be cooled by the stesa being generated by the covered portion of the core. The FOAM code F
as documented in topical report BAW-10064 was developed to calculate two-phase mixture heights during this time period.
From the CRAFT code, core liquid height, core power, and core pressure as functions of time are input into FOAM to calculate the swen level. The core liquid height is calculated from the inner vessel liquid I
volume, which is the sum of th~e liquid volume in the lower plenum, core, 1
core bypass, upper plenum, and upper head. For the CRAFT model used, i
this volume is the sum of the liquid volume in nodes 13 and 2.
The volume calculated is distributed in the inner vessel column so that the lower plenum will be filled with liquid. The remainder of the water is i
in the core and is used to calculate the quiet core liquid height. This approach is conservative since no credit is taken for the mixture that wi n exist in the lower plenum due to primary metal heat addition and flashing. This leads to less water in the' core than is expected and L
thereby results in lower swell levels, The FOAM code calculates the swell level based on a stesa bairres for which stesa lost from the mixture via bubble rise win be equal to steam production via core power. For this report, the average core heat rate is used The hot cham el would generate more steam and consequently l
would result in a higher swell level.
Cladding temperature transients for sman breaks will be more se-vere for axial power shapes peaked toward the top of the reactor because of the potential for steam cooling in the upper regions. Also, since 3-3 Babcock & Wilcox
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RESULTS OF ANALYSIS 4
2 This section presents a detailed evaluation of the breaks consid-ered, along with explanations of the phenomena involved. Size effects were analyzed at the pump discha'ge, and one case was analyzed at the r
pump suction to determine location effects.
)
4.1.
Explanation of curves The following explanations are provided to aid in understanding the i
t parameters illustrated in the curves.
Core mass flux - This is a plot of the smoothed average core mass flux as predicted by CRAFT and reduced by 20% for input into THETA.
I Core power - This curve indicates the normalized thermal power as calcu-lated by CRAFT.
Pressure - This is the pressure in node 2 as calculated by CRAFT.
1 Inner vessel liouid volume - This curve shows the total liquid volume of the inner vessel (including the lower plenum, core, and upper plenum, ex-ciuding the downcomer).
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Claddine temperature - This is the cladding temperature as calculated by THETA. The term " hot spot" means the highest single-point cladding tem-perature calculated for this accident.
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4.2.
Core Floodina Line Break l
Because of its unique location, a double-ended rupture of the core flooding line limits the ECC systems available to control the accident.
This break leaves one CFT, due HPI, and a portion of one LPI pump avail-able for core cooling. The LPI flow is available because it is cross-connected to each core flooding line. To ensure a conservative calcula-tion, no credit for LPI was taken until the system pressure fell below 165 psig. A 9-inch restrictor in the core flooding line nozzle limits the size of the break to 0.44 ft2, u
4-1 Babcock & Wilcox 4
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Because of the break size, the reactor coolant system depressurizes rapidly. Reactor trip and pump trip occur within the first 2 seconds.
Figures 4-10 and 4-11 show the pressure and power transients after the accident. The reactor coolant systes pressure drops to core flooding tank actuation pressure at 128 seconds. Due to the addition of subcooled water from the core flooding tanks, the reactor coolant system starts to depressurize faster at this point. Since the core flooding tanks depres-e surize the downconer initially, a flow reversal in the path conne. ting the downconer and lover. plenum causes an initial decrease in :le inner vessel liquid volume as shown in Figure 4-13.
However, once the down-comer head becomes large enough to overcome this effect, the inner vessel refills rapidly. At 176 seconds, the LPI pump starts -to provide makeup
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to the reactor vessel. This pumped injection becomes sufficient to pro-
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vide long-term cooling at 200 seconds. Figure 4-12 shows the leak flow for this break.
Figure 4-14 shows the results of the PGAM calculation. Again, this curve is different from that predicted by CRAFT for the same reasons men-tioned in section 4-1.
Figures 4-15 through 4-18 show pertinent param-eters for the temperature analysis. A peak cladding temperature of 1178F is reached at 180 seconds. Since this is a low temperature, no metal-water reaction or cladding swelling is expected.
4.4.
0.3-ft2 Break at Pumo Discharge c
The 0.3-ft2 break at pump discharge is considered as a medium-sized small break. Figures 4-19 through 4-22 are plots of the pressure tran-sient, power transient, leak flow, and inner vessel liquid volume for this j
case. Since the break is still large, there is a rapid depressurization l
causing reactor trip and pump trip, which initiatas pump coastdown, within the first 2.0 seconds. The pressure decays to 600 peig at 234 seconds and l
initiates core flooding tank flow. This action causes the inner vessel to refill with water. The pressure continues to drop at a slightly faster rate after actuation of the core flooding tanks. Injection from the LPI pump starts at 347 seconds. At 400 seconds, long-term cooling is estab-lished and the transient is carminated.
The mixture height is calculated to be above the top of the core at all times. No temperature calculation is necessary because heat trans-for within the mixture will be by a nucleate boiling process and the 4-3 Babcock siWilcox
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5 seconds of the accident. At 280 seconds, two-phase fluid flows through I
p the break and,results in a more rapid depressurization of the reactor
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coolant. system. For the break at the discharge, this effect occurred at 200 seconds. After this time, the break at the suction depressurizes faster than the break at the pump discharge due to higher quality fluid exiting the systesf The core flooding tank actuation pressure is reached at 775 seconds. At 1300 seconds, the LPI pump begins to inject water to 1*
the reactor vessel, resulting in an increase in the inner vessel inner F
volume. The auxiliary feedwater pump flow stops at 1390 seconds after M
the break occurs and causes an increase in the system pressure. Long-i term cooling is established at 2550 seconds. The core remains covered by a two-phase sixture throughout the transient, and no cladding tempera-ir ture increase will occur.
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As was shown, no major differences in response resulted from the change of location. The lower pressures that occur during the suction break result in increased pumped injection. Therefore, small breaks at the pump discharge are considered to be slightly worse than breaks at the pump suction.
4.7.
0.05-ft2 Break at Pump Discharme m
The 0.05-ft2 break at pump discharge is the smallest break analyzed.
2 break at This break responds basically in the same manner as the 0.1-ft the pump discharge as shown in Figures 4-31 through 4-34.
The reactor trip occurs 8 seconds after the break, the auxiliary feedwater pump shuts off at 1520 seconds, the core r===4na covered by a two-phase liquid throughout the accident, and the cladding temperature never exceeds its prebreak value. The results for this break show that the HPI pumps alone are sufficient to handle breaks of this size and smaller since the core flooding tanks inject water for only a short period of time and thus play an insignificant role -in controlling the accident.
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Normalized Power for Core Flooding Line Break i
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Inner Vessel Liquid Volume for Core Flooding Line Break 3500 3000 P
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Top of Core 3 1500
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Bottom of Core 500 i
0 0
100 200 300 400 Time, s 49 Babcock & Wilcox 4
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Hot Spot Cladding Temperature for Core Flooding Line Break 1800 P
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4-15 Babcock & Wilcox
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Leak Flow for 0.5-ft2 Break at Ptamp Discharge 12,000 L
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Core Mixture Height for 0.5-ft2 Break at Pump Discharge 12 r
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4-23 Babcock & Wilcox-t
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Inner Yassel Liquid Volume for 0.3-ft2 Break at Pump Discharge 3500 x
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Inner Yassel Liquid Volume for 0.1-fc2 Break at Pump Discharge 3500 l
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4-31 Babcock & Wilcox
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Inner Vessel Liquid Volume for 0.1-ft at Pump Suction L'
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REFERENCES 4
u 1 CRAFI-Description of Model for Equilibrium LOCA Analysis Program, BAU-10030, Babcock & Wilcox, Lynchburg, Virginia, October 1971.
2 B. M. Dunn, C. D. Morsac, and L. R. Cartin, & ltinode Analysis of Core Flooding Line Break for B&W's 2568-MWt Internals Vent Valve Plants, BAW-10064, Babcock & Wilcox, Lynchburg, Virginia, April 1973.
3 C. J. Hocevar and T. W. Wineinger, THETA 1-B - A Computer Code for Nuclear Beactor Core Thermal Analysis, IN-1445, Idaho Nuclear Corp.,
February 1971.
L 4
C. E. Parks, B. M. Dunn, and R' C. Jones, kitinode Analysis of Small j-Breaks for B&W's 2568-MWt Nuclear Plants, BAW-10052, Babcock & Wilcox, u.
Lynchburg, Virginia, September 1972.
J. F. Wilson, R. J. Grenda, and J. 7. Peterson, "The Velocity of
~
Rising Steam in a Bubbling 'Aro-Phase Mixture," ANS Transactions,.5_
(1962), p 151.
6 C. D. Morgan and H. S. Kao, TAFT - Fuel Pin Temperature and Gas Pres-i sure Ar.alysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, April 1972.
7 Three Mile Island Unit 1 Fuel Densification Report, BAW-1389, Babcock
& Wilcox, Lynchburg, Virginia, June 1973.
8 B. M. Dunn, et al., B&W's ECCS Evaluation Model, BAW-10104. Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, October 1975.
l l
l Babcock & Wilcox 5-1 i
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