ML20084T259
| ML20084T259 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/31/1995 |
| From: | Olson C, Laura Smith, Sung Y WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19330G001 | List: |
| References | |
| WCAP-14372, WCAP-14372-R, WCAP-14372-R00, NUDOCS 9506120360 | |
| Download: ML20084T259 (31) | |
Text
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L warrmanount NON-PROPRMARY class 3
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- WCAP-14372-1 t-i AP600 LOW FLOW -
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CRITICAL HEAT FLUX (CHF)
TEST DATA ANALYSIS I
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i C. A. Olson L D. Smith III l
Y. Sung j
I May 1995 l
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i WESTINGHOUSE ELECTRIC CORPORATION (A'm A)
P. O. Box 355 Pittsburgh, Pennsylvania 15230-355
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C 1995 Westinghouse Electric Corporation 9506120360 950531 All Rights Reserved PDR ADOCK 05200003 A
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TABLE OF CONTENTS 9
Section
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1.0.
INTRODUCTION 1-1
- 2.0 TEST FACILIT'I 2-1 3.0 TEST SECTIONS 3-1
' 4.0 TEST PROCEDURE 4-1 5,0 TEST PARAMETERS 5-1 6.0 DATA
SUMMARY
6-1 7.0 DATA ANALYSIS 7-1 8.0 ADJUSTMENT TO WRB-2 CORRELATION 8-1 a-
9.0 CONCLUSION
9-1
10.0 REFERENCES
10-1 r
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LIST OF TABLES l
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Tat!f Title fage I
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6-1 AP600 CHF Test Results 5x5 Typical Cell 6-2
' 6-2 AP600 CHF Test Results 5x5 Thimble Cell 6-4 l
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7-1 Summary of WESTAR Modeling 7-2 t
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LIST OT FIGURES h
Title fan 2-1 Elevation View of the Pressure Vessel and Test Section 2-2 3-1 Typical Cell Cross Section 3-2' 3-2 Thimble Cell Cross Section 3 3-3 Axial Geometry 3-4 3-4 Axial Power Profile for Westinghouse CHF Test 3-5 7-1 WRB 2 Measured-to-Predicted vs. Local Mass Flux 7-2 7-2 WRB-2 Measured-to-Predicted vs. Local Pressure 7-3 8-1 Adjusted WRB-2 Measured-to-Predicted vs. Local Mass Flux 8-3 8-2 Adjusted WRB-2 Measured-to-Predicted vs. Local Pressure 8-4 0
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1.0 INTRODUCTION
One of the design bases for pressurized water reactors (PWR), including future AP600 reactors, is to prevent the limiting fuel rod in the reactor core from reaching departure from nucleate boiling (DNB) for condition I and 11 events. The heat flux at DNB is often referred to as critical heat flux (CHF).
DNB correlations are developed from experimental data that simulate the reactor fuel design and core conditions. CHF is a function of local fluid conditions and can also be dependent upon fuel design.
'Ihe primary DNB correlation used for the analysis of the AP600 fuel is the WRB-2 correlation (Reference 1).
The applicable range of parameters for the WRB-2 correlation is:
Pressure 1440 s P s 2490 psia Local Mass Flow 0.9 s G s 3.7 lbm/hr-ft.2 Local Quality
-0.1 s Xw s 0.3 Heated Length, Inlet to DNB Location 1,514 feet Grid Spacing 10 s gsp 5 6 inches 2
Equivalent Hydraulic Diameter 0.37 s D, s 0.51 inch Equivalent Heated Diameter 0.46 s D, s 0.59 inch A complete loss-of-flow accident with all reactor coolant pumps (RCPs) tripped is a condition III event, but it is analyzed as a condition II event. A locked RCP rotor accident is a condition IV event.
In the locked rotor analysis, fuel rods having a DNB Ratio (DNBR) below a DNBR limit are assumed to fail for radiological assessment.
g In the postulated AP600 loss-of-flow and locked rotor accidents, die local mass fluxes in the hot channels at the DNB-limiting time steps are about 0.6 x 10'lbm/hr-ft.2 to 0.7 x 10'lbm/hr-ft.2 which are outside the applicable range of the WRB-2 correlation. To assess CHF performance of the AP600 fuel design at Nw-llow conditions, DNB tests were conducted at the Columbia University Heat Transfer Researce Facility in New York between July 1993 and February 1994.
'Ihis report provides a description of the DNB test facility, test sections, and testing procedure for data collection. 'Ihe report also describes data analysis and adjustment to the WRB-2 correlation based on the test results. The report qualifies the adjusted WRB-2 correlation for acceptance for die AP600 loss-of-flow and locked rotor applications.
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2.0 TEST FACILITY Westinghouse has obtained all of its CHF data from the Columbia University Heat Transfer Research Facility (HTRF). De facility consists of an instrumented high pressure loop that can supply water at pressures up to 2500 psia, flow rates up to 650 gpm, and inlet temperatures up to 650'F. The power supply is capable of delivering up to 12.5 megawatts dc. A description of the Columbia University test facility can be found in Reference 2.
The major components of the heat transfer loop are:
Circulating pumps Flow control and measuring spool piping section Test section housing Heat exchangers (HXs)
Mixing tec Water purification system Feedwater supply Makeup system Bleed system Figure 2-1 shows the test section housing. It has only five major components:
Pressure housing Grid plate Top adapter Shroud box Bottom adapter Water from the measuring spool pipe enters the top of the pressure housing, flows down between the annulus, formed by the shroud and the pressure housing inner wall, passes through the bottom adapter holes and turns upward into the flow channel containing the test section. De resulting steam water mixture flows through the top adapter and grid plate into the mixing tree.
I he instrumentation measurements used for the CHF tests include mass flux at'the test section inlet, water temperature and total pressure at the test section inlet and outlet, differential pressures between axial locations in the test section, temperature in different sections of the test loop, and total de power to the test section, and heater rod wall temperature. A computer-controlled data acquisition system recorded data during testing.
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e 3.0 TEST SECTIONS CHF test sections were constructed to accurately reflect the AP600 fuel design. There are two test sections: a typical section with heated rods only and a thimble test section containing a guide thimble tube in the center. The electrically heated rods were arranged in a 5x5 array with a 0.496.in.2 pitch and a heated length of 14 feet (which bounds the actual fuel heated length of 12 feet). The heated rod diameter is 0.374 inch, and the thimble tube diameter is 0.474 inch. Similar to previous CHF tests (Reference 1), the test sections have a non-uniform radial power distribution with the peripheral rods having lower power than the interior rods. Figures 3-1 and 3-2 show the test configurations and radial power distributions for the typical and thimble test sections, respectively.
The zircaloy mixing vane (MV) grids and the intermediate flow mixer (IFM) grids were placed alternatively in a 10 inch spacing along the rod buadles. Thermocouples were located in the rods at diffeient axial locations for CHF detection. Figure 3-3 shows the locations of grids and thermocouples along the heated length. A nonuniform axial power profile, similar to the power profiles used for previous CHF tests, was used for the AP600 tests (Reference 1). Figure 3-4 shows the axial power profile.
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- 4.0 TEST PROCEDURE he test procedure consisted of a cold-flow test, a heat balance test, and a CHF test.
De cold-f'ow test consisted of a series of pressure drop measurements on the rod bundle after the test section was installed in the test loop and before the power was connected. Differential pressure drops were taken at nominal conditions of 1000 psi.,80*F, and at several flow rates. Pressure drop data established a basis for comparison of hydraulic integrity of the test section.
De heat balance test included two heat balance checks, confirming that the test loop and test section instrumentation were in proper working order. The test was performed when the test loop reached 6
2 equilibrium after initial heat up. Nominal conditions of 1500 psi,400*F,1.5x10 lbm/hr-ft mass flux and 1.2x10' watt total power to the test section were set for heat loss calculation. Acceptance criterion for heat loss is less than 3 percent of the total heat input to the test section.
De CHF test was performed by maintaining constant test section outlet pressure, inlet temperature, and mass flux. Total power to the test section was then increased in small increments (less than 30 kW per step) until a temperature excursion occurred in one or more of the thermocouples positioned in the heater rods. De excursion was about 30'F, varying with system conditions. When the temperature excursion occurred, power to the test section was then reduced and preparation for the next test condition began.
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5.0 TEST PARAMETERS '
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'The range of system conditions for low-flow testing are:
Inlet Temperature (*F) 300 - 536 Inlet Mass Flux (10 lbm/hr ft.')
0.5-1.0 5
Exit Pressure (psia) 1500 - 2400 The system conditions bound the low-flow conditions encountered in the AP600 loss-of-flow and locked rotor accident analyses. 'Ihe local mass fluxes in the hot channel at the DNB-limiting time.
increments are about 0.6 x 10' to 0.7 x 1& lbm/hr.-ft.2 As a result, the range ofinlet mass flow for the testing was 0.5 x 10' to 1.0 x 10' Ibm /hr-ft.2, 1
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SUMMARY
The low-flow CHF data are presented in Tables 6-1 and 6-2.~ These tables give inlet pressure, inlet mass flux, inlet temperature, and average bundle heat flux. They also identify the thermocouples that indicated a CHF event.~ Average heat flux includes a small adjustment to account for power losses in electrodes at the ends of each rod. The adjustment includes the changes in rod and electrode electrical resistivity as a function of temperature.
Rod numbering begins from one of the corner rods and continues around the 5 x 5 rectangular array in a concentric spiral, ending with the center rod. The numbering scheme used for the CHF rods is XX.Y. where XX is the rod number and Y is the thermocouple number. For example, CHF rod 24.2 indicates that CHF was observed in the second thermocouple of the 24* rod.
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i TABLE 61 AP600 CHF TEST RESULTS 5x5 TYPICAL CELL Inlet Idass Average Inlet Flux Inlet -
Heat Flux Pressure (Ibadr ft.')
Temperature (Btu /br-ft.2).
Thermocouples Indicating (b, )
g Run
-(psia)
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(*F) le Clip')
6030 6031 6032 6033 6034 6035 6036 6037 6038 6039 6040 6041 6042 6043 6044 6045 6046 6047 6048 6049 6050 6051 6052 6053 6054 6055 u:\\np600\\20llw.non:Ib-Om195 62 REVISION: 0
l TABLE 6-1 (Cont.)
AP600 CHF TEST RESULTS 5x5 TYPICAL CELL Inlet Mass Average Inlet Flux Inlet Hes.t Flux Pressure (Ibm /br ft.')
Temperature (Btu /hr ft.2)
Thermocouples Indicating Run
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('F) 10' CHI *
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6056 6057 6058 6059 6060 606]
6062 6063 7021 7022 7023 8016 8017 8018 6019 8026 8027 2028 8029 8030 8031 8032 8033 8014 8035 8036 8037 8038 8039 8040 k041 8042 Note:
(1) Thennocouple identifica6on example: 24.2 = rod #24, axial posidon #2.
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TABLE 6-2 AP600 CHF TEST RESULTS 5x5 THIMBLE CELL Inlet Mass Average Inlet Flux Inlet Heat Flux Pressure Obm/br.ft.2)
Temperature (Btu /hr ft.2)
Thermocouples Indicating Run (psia) 10' (F) 10' CHP" (b,c) 2022
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a TABLE 6.2 (Cont.)
AP600 CIIF TEST RESULTS 5x5 TIIIMBLE CELL Inlet Mass Aserage Inlet Flux Inlet Ileat Flux Pressure (thm/hr-tt.2)
Temperature (Btu /hr-ft.')
Thermocouples Indicating Run (psia) 10' (F) 10' CEII5" 2051
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2052 2053 2054 2055 2056 2057 2062 2063 2064 2065 2066 2067 2068 2064 2070 2071 2072 2073 2074 2075 2076 2077 207k 2079 2080 2081 2082 2081 Note:
(1) "Ihermocouple identification example: 24.2 = rod # 24, axial position #2.
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7.0 DATA ANALYSIS The WRB-2 correlation is the primary DNB correlation used for the analysis of AP600 fuel. The original data range of the local mass flux for the WRB-2 correlation is between 0.9 and 3.7 x 10' lbm/hr-ft.2 (Reference 1). In the postulated loss-of-flow and locked rotor accidents, local mass fluxes in the hot channels at the DNB-limiting time steps are less than 0.9 x 10'lbm/hr-ft.2 The CHF data described in Section 6.0 were used to evaluate the WRB-2 correlation for the loss-of-flow -
and locked rotor applications.
' The WESTAR code (Reference 3) was used to calculate local fluid conditions in subchannels for each data point in Tables 6-1 and 6-2. WESTAR is a subchannel analysis code that calculates three-dimensional flow and enthalpy distributions in rod bundle geometries. The WRB-2 correlation predicts CHF based on local fluid conditions from WESTAR. WESTAR has been accepted by the U.S. Nuclear Regulatory Commission (NRC) for licensing reactor core thermal-hydraulic calculations, including the use of the currently licensed WRB-2 DNBR limit of 1.17 (Reference 4).
'Ihe WESTAR test section modeling is consistent with the M. STAR model for the AP600 reactor core. A higher turbulent mixing coefficient is used for data analysis but for AP600 safety analysis, the turbulent mixing coefficient is set at a lower value. Table 7.1 summarizes the WESTAR modeling of test sections.
Comparison of the predicted CHF with the measured CHF showed that the WRB-2 correlation tends to overpredict CHF at low-flow conditions. 'Ihe magnitude of overprediction depends greatly on local mass flux and slightly on local pressure. Figures 7-1 and 7-2 show the ratio of measured to predicted CHF (M/P) plotted against local mass flux and local pressure, respectively. An adjustment to the WRB 2 correlation is necessary to correct the CHF overprediction.
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TABLE 7.1
SUMMARY
OF WESTAR MODELING Radial geometry 1/4th test section Axial modal length, inch 4.2 Mixing coefficient PA3 = 2.05' Mixing vane grid K K thimble 0.92 K-typical 0.84 K-side 0.82 K-comer 0.84 IFM grid K K-thimble 0.75 K-typical 0.60 K-side 0.56 K-comer 0.55 Note:
The mixing coefficient used for safety analyses is PA3 = 1.32 (PA3 = 2.05 is equivalent to thermal f
diffusion coefficient (TDC) = 0.59 and PA3 = 1.32 is equivalent to TDC =.038 used in THINC code)
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I 8.0 ~ ADJUSTMENT TO WRB-2 CORRELATION To correct the CHF over-prediction, a multiplier is applied to the WRB-2 predicted CHF as follows:
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where:
_ q"rms.w, s = adjusted _WRB 2 predicted CHF (as) local mass flux,10'lbm/hr-ft.2 gioc
=
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Jtax) local pressure, psia *10' ploc
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The adjusted WRB-2 M/P's are plotted in Figures 8-1 and 8-2 agairist local mass flux and local pressure, respectively. De adjusted WRB-2 predictions are conservative compared to the low-flow data. The statistics of the adjusted M/P's for CHF data in Tables t>-1 and 6-2 are listed below:
Table #
- of Data M/P Mean*
M/P Std Dev.*
6-1 58
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6-2 58
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6-1 & 6-2 116
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- Adjusted WRB-2 he data range for the adjusted WRB-2 correlation is summarized below:
1503 5 Pressure 5 2430 psia 0.48 s Local Mass Flow 5 1.0410 lbm/Ir-ft.2 6
0.0 s Local Quality 5 0.81 6
in AP600 safety analyses, if local mass flux in the hot channel is between 0.48 x 10 and 1.(M x 10 lbm/hr-ft.2, the adjusted WRB-2 correlation is used for DNB Ratio (DNBR) 6 l
calculations. The WRB-2 correlation is used when the local mass flux is between 1.0 x 10 6
and 3.7 x 10'lbm/hr-ft.2 1
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p Consistent with design acceptance criterion in the U.S.~ NRC Standard Review Plan NUREG-0800 (Reference 5), the DNB design criterion is that there is no DNB occurrence with a 95 percent probability at a 95 percent confidence level (95/95) on the most limidng fuel rod during normal operation, anticipated operational transients, and any transient conditions arising from faults of moderate frequency (condition I and 11 events). The loss-of-flow accident with all coolant pumps tripped is a condition ill event, but it is analyzed as a condition 11 event. De locked rotor accident is a condition IV event. In the locked rotor analysis, fuel rods having DNBRs below a DNBR limit are assumed to fail for radiological assessment.
To meet the 95/95 design criterion, a limiting value of DNBR based on the correlation statistics can be determined by the Owen's method (Reference 6). Owen prepared tables which give values of k, such that at least a proportion of p of the population is greater than (M/P)4va - k,*s with confidence u, where (M/P),vo is the sample mean and s is standard deviation. Owen's factor k, accounts for uncertainties in the size of data sample.
De 95/95 DNBR limit for the adjusted WRB-2 correlation at the low-flow conditions is calculated based on the M/P statistics of DNB data in Table 6-2 only. As compared to the statistics based on Tables 6-1 and 6-2 or combined Tables 6-1 and 6-2, the M/P statistics based on Table 6-2 yields the highest DNBR limit. The DNBR limit to be applied to the AP600 loss-of-flow and locked rotor analyses is [
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95/95 DNBR Limit
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9.0 CONCLUSION
To assess the CHF performance of the AP600 fuel design at the low-flow conditions DNB tests were conducted at the Columbia University Heat Transfer Research Facility using the test configurations reficcting the AP600 fuel design. The test conditions bounded the low-flow conditions encountered in the AP600 loss-of-flow and locked rotor accident analyses. The DNB test data were analyzed using the WESTAR code and the WRB-2 correlation. Based on the comparison with the test results, an adjustment was made to the WRB-2 correlation. The 95/95 DNBR limit with the adjusted WRB-2 correlation is [
f"' for the AP600 loss-of-flow and Mcked rotor applications.
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10.0 REFERENCES
1.
Davidson, S. L. (Editor), Reference Core Report - VANTAGE 5 Fuel Assembly, WCAP-10444-P-A, September 1985, " VANTAGE SH Fuel Assembly", WCAP-10444-P-A, o
Addendum 2-A, April 1988.
2.
Hill, K. W., F. E. Motley, F. F. Cadek, and A. H. Wenzel, Effect of 17x17 Fuel Assembly Geometry on DNB, WCAP-8296-P-A, Febniary 1975.
3.
Ho, S. A., C. A. Olson, and I. K. Paik, WESTAR: An Advanced Three-Dimensional Program for the Thermal-Hydraulic Analysis of Light Water Reactor Cores, WCAP-10951-P-A, June 1988.
4.
Letter from A. C. Thadani (NRC) to W. J. Johnson (W), Acceptancefor Referencing of Licensing Topical Report WCAP-10951, WESTAR: An Advanced Three-Dimensional Povgram for Thermal-Hydraulic Analysis of Light Water Reactor Cores, June 20,1988.
5.
NUREG-800, Standard Review Plan, Section 4.4, Thenna! and Hydraulic Design, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Revision 1, July 1981.
6.
Owen, D. B., Facwrsfor One-Sided Tolerance Limits andfor Variables Sampling Plants, Sandia Corporation, SCR-607, March,1960.
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