ML20247G862
| ML20247G862 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 03/30/1989 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20247G537 | List:
|
| References | |
| NUDOCS 8904040334 | |
| Download: ML20247G862 (18) | |
Text
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CESSAR EnnflCATION O
APPENDIX 4A SYSTEM 80 REACTOR FLOW MODEL TEST PROGRAM O
O grog saqr m
CESSAR Ennncaucu O
EFFECTIVE PAGE LISTING CHAPTER 4 A.PENDIX P 4A T3ble of Contents Page Amendment i
11 111 Text Pace 4A-1 4A-2 4A-3 4 ?.-4 j Tables _ l 4A-1 i l
Fiqures l
4A-1 4A-2 4A-3 4A-4 4A-5 4A-6 4A-7 I
O Amendment B March 31, 1988
CESSAR 82&"ic.1 cn O i TABLE'OF CONTENTS CHAPTER 4 1
APPENDIX 4A f 1
Section Subiect Pace No.
1.O INTRODUCTION 4A-1
-l 2.O -DESCRIPTION OF FLOW-MODEL- 4A-1 .
2.1 PRESSURE VESSEL AND CORE SUPPORT STRUCTURES. 4A-1 2.2 MODEL CORE '4A-1 2.3' MODEL INSTRUMENTATION 4A-1 i
3.0 DESCRIPTION
OF TEST FACILITY 4A-2 3.1 TEST FACILITY AND OPERATING CONDITIONS 4A-2 l O 3.2 TEST LOOP 4A-2 3.3 DATA ACQUISITION SYSTEM 4A-2 3.4 CALIBRATION STANDARDS 4A-2 4.0 DATA ANALYSIS 4A-2 4.1 MODEL POINT PRESSURES 4A-2 4.2 CORE INLET FLOW DISTRIBUTION 4A-3 4.3 _ CORE OUTLET PRESSURE DISTRIBUTION 4A-3 4.4 REliCTOR VESSEL PRESSURE DROP 4A-4 4.5 COMPONENT HYDRAULIC LOADING 4A-4 i
1
, i l
- 1. 1
. 4 CESSAR nuincuion O
LIST OF TABLES CHAPTER 4 APPENDIX 4A l
Table subiect 4A-1 Reactor Vessel Best Estimate Loss Coefficients &
Pressure Drops I
O t
O ,
i i
11
CESSAR 88Hncamn O
LIST OF FIGURES CHAPTER 4 APPENDIX 4A Ficure Bubiect 4A-1 Reactor Flow Model
\ 4A-2 Comparison of Reactor and Model Fuel Assembly Layout 4A-3 Pressure Tap Locations in the Reactor Flow Model 4A-4 Test Loop Schematic 1
4A-5 Schematic of Data Acquisition System 4A-6 Core Inlet Flow Distribution Qg d 4A-7 Core Exit Euler Numbers, E*
9 1
i l
O iii
CESSAR nuirlCATION O
APPENDIX 4A - SYSTEM 80 REACTOR FLOW MODEL TEST PROGRAM
1.0 INTRODUCTION
Flow model tests have been conducted to determine the hydraulic performance of the System 80 class reactors. Tests in 1973-1975 at Kraftwerk Union (KWU) with a 1/8 scale model of the downcomer and lower support structure, and in 1975-1976 at C-E with a 3/16 scale model of the outlet plenum region, were conducted to refine basic hydraulic design. Objectives were to produce an acceptable core inlet flow distribution, and to minimize hydraulic loadings and flow resistance. The reactor core was not represented in these tests, except for simplified provisions to represent a portion of the core inlet resistance for inlet flow distribution mapping in the KWU lower plenum tests. The test results obtained through 1976 were treated as preliminary in view of the simplified core modelling. Results were used 4.n design analyses of core thermal margin, system pressure drop, and component hydraulic loading.
Final verification of reactor design hydraulic parameters is based on tests with a 3/16 scale model having all structures in the main flow paths, including a dynamically scaled core. This 9 model was constructed in 1976-1977 and tested in 1978, in the C-E Nuclear Laboratories. These final tests are the subject of the further discussion in this Appendix.
2.0 DESCRIPTION
OF FLOW MODEL 2.1 PRESSURE VESSEL AND CORE SUPPORT STRUCTURE _S A cross-sectional view of the flow model is presented in Figure 4A-1. Geometric similarity to the reactor main flow paths is maintained, with a scaling factor of 3/16, except in the model core. Relatively stagnant regions at the top of the downcomer and in the upper guide structure and closure head volume are truncated for ease and economy of model assembly.
2.2 MODEL CORE The model core consists of an array of 241 square tubes, each representing one fuel assembly. Six levels of flow resistor plates match the axial flow resistance of reactor fuel assemblies, and approximately match the axial distribution of axial flow resistance over the length of the assembly. Aligned round holes through adjoining tube walls match the resistance to cross-flow between reactor fuel assemblies. Model and reactor fuel assemblies are compared in Figure 4A-2. The model core 9 design and associated "open-core" flow model testing technique follow the methodology of flow model tests for the C-E 3400-Series reactors, as described in CENPD-206-P(1).
CESSAR 8Hn,"lCATION O
2.3 MODEL INSTRUMENTATION Model instrumentation consists of wall static pressure taps in the inlet and outlet ducts, at the top and bottom of the downcomer, on the flow skirt and bottom plate, in the inlet and outlet of each core tube, at several points on the upper guide structure, and in the closure head volume. These taps provide for assessment of the breakdown of reactor vessel pressure drop, component steady state hydraulic loading, and for measurement of core inlet and core outlet pressure boundary conditions. A detailed summary of pressure tap locations is provided in Figure 4A-3.
3.0 DESCRIPTION
OF TEST FACILITY l l
3.1 TEST FACILITY AND OPERATING CONDITIONS Testing was conducted in the C-E Large Scale Hydraulic Test Facility, TF-15. For the configuration represenP.ing full flow with four operating reactor coolant pumps, model flow was set at 11,000 gpm. All tests were conducted at approximately 80*F fluid temperature. At these conditions, flow in the model ig fully turbulent, with an outlet duct Reynolds Number of 2.6 x 10 .
Tge corresponding reactor Reynolds Number at full power is 1.5 x 10 ,
It is expected that non-dimensionalized pressure drons and flow distributions do not chance at the hiaher reactor Reynolds Numbers.
3.2 7EST LOOP The TF-15 test loop, as used for System 80 Flow Model full flow testing, is depicted in Figure 4A-4. Three circulating pumps are required to provide the 11,000 gpm model flow rate. Individual inlet and outlet duct flow settings are established with flow control valves and calibrated flow meters. Flow meter signals are continuously fed to the data acquisition system to verify consistency of flow settings.
3.3 DATA ACQUISITION SYSTEM The data acquisition system used for this test is depicted in F!gure 4A-5.. Model point pressures are sequentially connected through computer-operated solenoid valves to a series of differential pressure transducers. Point pressures are read against an internal reference pressure at mid-elevation in the model core. Electrical output is repeatedly scanned and averaged for each point after a dwell period in which pressure differences are allowed to stabilize. Switching and digital voltmeter readout are controlled by a conputer. Operator control of test progress and screening of measurements are accomplished with a teletypewriter. Data are recorded on magnetic tape for further reduction by a digital computer.
CESSAR unifium.
O 1 3.4 CALIBRATION STANDARDS Calibration of the instruments is made utilizing fixed water columns or variable height mercury columns, as appropriate for the range - of each instrument. Calibrations are made at the 1 beginning of each test run, and are confirmed upon completion of I each test run.
4.0 DATA. ANALYSIS 4.1 MODEL POINT PRESSURES Model point pressures, measured relative to an internal reference pressure at core mid-elevation, are converted to Euler numbers of one of several forms,
- a. For the planar pressure distributions at the core inlet and core outlet, Pin and Pout:
E i == (Pi - Pi) / (Pin - Pout) )
l b. For other points and spatially' averaged pressures:
E i := (Pi - Pin) / VH ref where VH ref is a reference model inlet velocity head.
- c. For point-to-point pressure differentials:
1 El == (P upstream - P downstream) / VH ref Euler numbers are readily converted to desired pressure drop loss coefficient and hydraulic loading coefficient forms, considering averaged data from repeated runs.
4.2 CORE INLET FLOW DISTRIBUTION l The System 80 core inlet flow distribution for the normal i condition with four operating reactor coolant pumps is provided in Figure 4A-6. At each fuel assembly location, the inlet flow is expressed as a fraction of the average fuel assembly flow rate in the core. Flow model test data, in the form of the core inlet pressure distribution, are scaled to reactor conditions and used in a TORC /HERMITE simulation of the System 80 core to determine the core inlet flow field. This technique for determining the core inlet flow distribution is discussed further in CENPD-206-P(l), Section 3.1.1.
CESSARE! Enc-O The uncertainty on the core inlet flow distribution includes the test measurement uncertainty and the uncertainty in the TORC /HERMITE calculation of the flow distribution from the measured pressure distribution. The typical uncertainty on the core inlet flow distribution (oQg /Q) is 0.06 at the lo level.
4.3 CORE OUTLET PRESSURE DISTRIBUTION The reactor core outlet pressure distribution is p.rovided in Figure 4A-7, for the normal condition with four operati.ng reactor coolant pumps. Euler numbers at fuel assembly locations express the core outlet pressure distribution in a non-dimensional form which is defined as, Py - P outlet i) outlet " -
core 9 core where: Pg = local static pressure, core outlet P = core average static pressure, core outlet K
core
= core overall loss coefficient, based on O core flow area q = averag core outlet velocity head, based on core core flow area The core outlet pressure distribution is obtained as a result of interfacing two analytical simulations: the first simulation is a representation of the core region, using the TORC code; the second is a multi-flow-path simulation of the upper plenum region between the core exit and the outlet nozzles. Input to the TORC '
code contains the core inlet flow distribution as determined earlier from flow model test data. Input to the upper plenum j simulation contains the flow resistances found in flow model i tests for this region. Matching of the interface conditions between the two simulations provides the core outlet flow and pressure distributions.
1 Uncertainty in the core outlet pressure distribution takes into )
account the uncertainty in the TORC representation of the core i and the uncertainty in the analytical model of the outlet plenum i and core exit regions. The typical uncertainty in the core {
outlet pressure distribution (AEg ) is 0.008 at he la level.
O l I 4A-4
CESSAR En@icavi:n O
4.4 REACTOR VESSEL PRESSURE DROP The System 80 reactor vessel incremental loss coefficients and pressure drops, based on test results, are provided in Table 4A-1. Data for the fuel assembly come from full scale flow tests on a typical fuel assembly. Frictional pressure drops for'the fuel assembly and the downcomer are based on standard friction factor methods. The remaining pressure drops are based upon results from the 3/16 scale flow model test.
Uncertainty in incremental loss coefficients from the full scale flow tests on the fuel assembly is considered to be independent of the uncertainty in the loss coefficients from the 3/16 scale flow model test. The uncertainties are combined by the root-sum-square technique.
The uncertainty in the total reactor vessel pressure drop due to test measurement uncertainty is calculated to be 4.4% at the la level.
4.5 COMPONENT HYDRAULIC LOADING
,A Reactor internal component design steady state hydraulic loads
's which are verified using scale model flow test data include the following:
- a. Core support barrel and upper guide structure uplift forces.
- b. Differential pressure loadings on the:
o Flow skirt o Bottom plate o Fuel alignment plate o Upper guide structure support barrel
- c. Steady state drag loading on the cylindrical shroud tubes in the outlet plenum and instrument nozzles in the lower plenum.
( Design values for these hydraulic loads, based on earlier flow model test results and on analytical methods,_are in all cases -
shown to be conservative on the basis of final model test results.
O f
4A-5 i
C E S S A R !a nnca m n
- TABLE 4A-1 REACTOR VESSEL BEST ESTIMATE LOSS COEFFICIENTS & PRESSURE DROPS Flow Path Stations Loss Coeff.. Ki Press. Drop. osi Temp. 'F Inlet Nozzle & 90* Turn 1-5 0.69 8.7 565 Downcomer, Lower Plenum
& Lower Supp. Structure 5-15 1.21 15.4 565 Fuel Assembly 15-20 1.27 15.9 595 l
Fuel Assy. Outlet to Outlet Nozzle 20-24 1.18 16.7- 624 Total Pressure Drop 56.7 O
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