ML20028E937
| ML20028E937 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/30/1982 |
| From: | Coffield R, Luggy F, Tang Y WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20028E932 | List: |
| References | |
| NUDOCS 8301280282 | |
| Download: ML20028E937 (68) | |
Text
{{#Wiki_filter:.- _ - CRBRF-ARD-0315 C' l Clinch River Breeder Reactor Plant VERIFICATION OF F@RE-2M COMPUTER CODE PART 11. Comparison with EBR-ll Natural Circulation Experiments Y. S. Tang ; 4 R. D. Coffield F. C. Luffy Prepared for the United States Depart-ment of Energy under Contracts DE-AC15-76CLO2395 and EW-76-C 0003. Any Further Distribution by any Holder of this Document or of the Data Therein to Third Parties Representing Foreign Interest, Foreign Governments, Foreign Companies and Foreign Subsidiaries or Foreign Divisions of U.S. Companies Should be Coordinated with the Director, Division of Reactor Research and Tech-nology, United States Department of Energy. l g W Westinghouse Electric Corporation ADVANCED REACTORS DIVISION C7FJ BOX 158 MADISON, PENNSYLV ANIA 15663 D 0 0 0 37 A PDR
O VERIFICATION OF F7RE-2M COMPUTER CODE, PART II COMPARISON WITH EBR-II NATURAL CIRCULATION EXPERIMENTS Y. S. Tang R. D. Coffield O F. C. Luffey NOV 1982 f APPROVED: R. A. Markley, Manager ' Thermal & Huids Systems Engrr. Westinghouse Advanced Reactors Division P. O. Box 158 Madison, Pennsylvania 15663 O
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TABLE OF CONTENTS , .g a - w.. , , s,
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- 1. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . ... , 1
- 2. DESCRIPTION OF EBR-II XX07 AND XX08 ASSEMBLIES. . . .. ... . .' i l 'j.[
2.1 ASSEMBLY XXO7 AND INITIAL CONDITIONS OF TEST F . . . . . l' ' 2.2 ASSEMBLY XX08 AND INITIAL CONDITIONS OF TEST 7A. . 1 k' ,.5
- 3. ANALYTICAL MODEL IN FORE-2M . . . . . . . . . . . . . . .'. . S.
- 4. FORE-2M ANALYSES. . . . . . . . . . . . . . . . . . . . . .,. 10 '
4.1 XX07 TEST F ANALYSIS , . . . . . . . . . . . . . . . ,. . 10 - 4.2 XX08 TEST 7A ANALYSIS. . . . . . . . .'. . . . . . . . . 8 l
- 5. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . '
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- 6. REFERENCES. . . ... . . . . . . . . . . . . . . . . . . . . . 25 '"
I APPENDIX A - COMPARIS0N OF FORE-2M NATURAL CIRCULATION ' - TRANSIENT PREDICTIONS WITH EBR-II, XX07 ASSEMBLY,
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APPENDIX B - COMPARISON OF FORE-2M NATURAL CIRCULATION RESULTS WITH EBR-II, XXO8 ASSEMBLY, TEST 7A DATA " e . g. Q s
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f \ O 1. INTRODUCTION For the LMFBR Natural Circulation Verification Program (NCVP) EBR-II natural circulation experiments are cons edered as an important source of data (Ref.1) for code verification. A significant amount of plant and reactor data have been collected during the residence of instrumented assemblies, XX07 and XX08, in EBR-II. As a part of the verification of F5RE-2M computer code, this report presents the comparison between FORE-2M analysis and selected data from EBR-II natural circulation ey9eriments. A summary of the type of plant transients for which data are available is shown in Table 1 (Ref. 2). Several different categories of transients are experienced with regard to primary and secondary flow variations and initial power vid primary flow conditions. Two set.s of tests were selected based on constant secondary system flow arid reasonable initial (pre-tripped) flow and power conditions. These are Test F of assembly XLO7 and Test 7A of assembly XX08. , Test F froa the XXO7 series involved a loss of primary forced flow which was beir:9 provided by the auxiliary pump while the reactor was shut down and the fission-product decay power was 1.6% of rated power (60 MWt) (Ref. 2). The seccacary flow wa, beld constant at 2% of its rated value. Among tha ficw and temperature measurements through the instrumented fueled driver assembly, coolart temperatures at inlet, mid-core and near-core exit were used to compare with FME-2M code predicted values. Test 7A from XA08 series was initiated under steady state operating conditions of Oy power at 17.1 MWt (28.5% rated power) and a flow rate of 2626 gpm (32.1% of - rated flow). The transient was initiated by interrupting the electrical power supply to the motor-generator set driving the primary pumps, the primary auxiliary pump having been previously de-energized. The secondary forced flow was maintained during the initial three minutes of the transient. Comparisons with code predicted coolant temperatures were made through "near center" elements (average of four fuel elements near the assembly center). Comparisons were also made with COBRA-WC code predictions (Ref. 3). Section 2 describes the design of EBR-II XX07 and XX08 assemblies and instru-mentations. Section 3 presents the analytical model of FORE-2M. The code , predicted values are compared with test data in Section 4. Conclusions are summarized in Section 5. The detail FORE-EM calculations for the temperature predictions of XXO7 assembly Test F data and XXO8 assembly Test 7A data are attached as Appendices A and B, respectively.
- 2. DESCRIPTION OF EBR-II XX07 AND XXO8 ASSEMBLIES 2.1 Assembly XX07 and Initial Conditions of Test F
' The XX07 assembly is an instrumented fueled assembly located in a converted control rod position. It contains 61 elements (outer diameter of 0.174 in.),
which were spiral wrapped with 0.049 in, helical spacer wire (lead = 6.0 in.). The fuel pellet is 0.144 in. OD and 13.5 in. long which is sodium bonded to the stainless steel cladding. The assembly configuration and shape is shown
- i. in Figure 1. The detailed dimensions are given in Section 3 of Appendix A.
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TABLE 1 (froin Ref. 2) 's
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EBR-II WHOLE CORE /WHOLE PLANT TESTS AND EENTS ' STEADY-STATE TEST IDENTIFICATION SECONDARY SYSTEM OR TRANSIENT FLOW POWER FLOW CONDITMNs EX-109 (XXO7) Steady-state A1: pumps off 0.11% det:y heating 0.4 to 6.1%, centro 11ed Test G (XX07) Steady-state All pumps off 0.5 to 1.3% fission 2.5 to 6.0%, ctntrolled heating Test F (XXO7) Transient Primary pumps off, pri. 1.6% decay power 2.0%, controlled aux. pump tripped Cover Lift Test (XX07) Transient Pri . puinps off, pri . aux. 0.58% decay power s 1%, controlled' pump on, reactor cover m - lifted to fuel handic LOF/ Scram of 1/1/75 (XXO7) Transient' Loss of power to pump 1 Scram from 100% Trip from 100; flow manual trip of pump 2, power on low flow with scram from 100% flow. Primary signal aux. pump on. Test lA (XX08) Transient Primary pumps off, pri. 0.15% decay power 3.3 to 10.5%, ,cn-aux pump tripped (XX08unirradiated, trolled; also tatural with 0.03% power) convective flo< Tests 1B-lE(XX08)' Transient Primary pumps off, pri. 0.17 to 0.63% decay - 2.6 to 9.0%, c'.ntrolled, aux. pump tripped power also natural cmvec-tive flow Test 2 (XX08) Transient Primary pumps off, pri. < 0.1% decay powea. Trip of .sec. puno from aux. pump on Plant isothermal tt 30, 50, 100L 1 ow at 580*F Test 7_A_(XX08L Transient Primary pumps tripped Scram from 29% Constant at 3M from 34% flo ri. aux. power pump off ,
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EBR-II WHOLE CORE /WHOLE PLANT TESTS AND EVENTS (CONTD) , I PRIMARY SYSTEM CONDITIONS - STEADY-STATE TEST IDENTIFICATION OR TRANSIENT SECONDARY SYSTEM FLOW POWER FLOW C0r.01110NS Test 7C (XX08) . Transient Primary pumps tripped 'from Scram from 36% Constant at 40% 40% flow, pri. aux. pump
- power off Test 8 (XX08) Transient Primary pumps tripped from Scram from 36% Secondary pump tripped
,40% flow, pri . aux. pump . from 40% flow 32 s off i after primary pump trip Test 8A (XX08) . Transient Primary pumps tripped from Scram from 36% Secondary pump tripped 40% flow, pri. aux. pump power from 40% flow 17 s j .off
! , it . r. , , g after primary pump trip Test 10 (XX08) Transient Primary pumps tripped from Scram from 100%
34% and 100% flow, pri. Sec. flow controlled at power preceding LOF 9%, nat. circ., or loss aux. pump off by 45 to 215 minutes of flow throttling LOF/ Scram of 10/1/77 (XX08) Transien't Primary pumps tripped from Scram from 100% Sec. Pump trip from 100% flow, pri . aux. pump power 100% flow with scram on LOF/ Scram of 11/2/77 (XX08) Transient Primary pumps tripped from Scram from 100% Sec. pump trip from
,100% flow, pri. aux. pump power 100% flow with scram on LOF/ Scram of 1/10/78 (XX08) 'Trandi,ent Primary pumps tripped from Scram from 1005 Sec. pump trip from 100% flow, pri. aux. pump power 100% flow with scram on .
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(0.6251D) FUEL INLET ELEMENT Figure 1. EllR-Il Emironmental Instrumented Subassembly XX07 8021-28 4
I A cross-section of the assembly illustrating the location and type of instru- l mentation used is shown in Figure 2. The assembly instrumentation included two inlet flow meters, ten fuel centerline thermocouples, three inlet and two outlet coolant thermocouples, three core midplane coolant thermocouples, and five coolant thermocouples near the top of the core. Test F of XX07 series was initiated after an anticipated shutdown scheduled for the end of reactor run 73 (completed 2099 mwd operation for about 34 days). The primary coolant pumps were maintained at full flow (a total bundle flow of 22,060 lb/hr at 700 F inlet temperature) for 5 min. following the reactor was scrammed, before they were tripped. The resultant flow dropped to approxi-mately 5% of the rated flow by the auxiliary pump. The latter pu:"p was tripped at about 24 min after the reactor scram and all primary system flow was then due to natural circulation. 2.2 Assembly XX08 and Initial Conditions of Test 7A The configuration of this assembly is different from XX07 in the use of flow mixer near the coolant outlet and the location of the coolant outlet thermo-couple within the assembly, as shown in Figure 3. The fuel pellet size, clad-ding thickness and sodium bond thickness in assembly XX08 elements are also slightly different (see Section 3 of Appendix B). The fuel bundle instrumen-tation loading plan is shown in Figure 4. Comparing with XX07 assembly, additional T/C were placed at 5.4 in., 12.7 in., 20 in and 36.7 in. levels above the core bottom level. b d Preceding the Test 7A, EBR-II was maintained at steady state (Run 97A) with the reactor power at 17.1 MW (28.5% of rated power) and flow at 2700 gpm (33% ) of rated flow). These conditions were held for about 3 hours until the auxiliary pump was turned off. For 9 minutes the primary flow was at 2621 gpm (32.1% of rated flow) and then the primary pumps were tripped and about 1.8-2.3 seconds after the pump trip, a low flow trip caused a reactor scram. The entire flow in the primary system was the result of natural circulation, following the initial flow coastdown. .The overall time period of interest for natural circulation analysis is the first two minutes after the primary pump trip (time = 0 at this point). The pretest flow condition in the XX08 assembly was: 665 F inlet temperature, 5804 lb/hr bundle flow and 110.1 kw assembly power. The relative radial and axial power profiles are given in Section 4 of Appendix B.
- 3. ANALYTICAL MODEL IN FORE-2M The FORE-2M computer code (Refs. 4&9) used a single channel model to represent either the average assembly condition or the average of local group of subchannels. For analysis of EBR-II XX07 assembly test data, the model used is shown in Figure 5. The fuel element is divided into seven axial and five radial nodal networks. The radial nodalization consists of three equal volume fuel nodes, one cladding node with sodium bonded gap and a coolant node. Six axial nodes are used to model the active core region and one axial node is used to represent the steel rod including the coolant entry region (0.5in.). The wire wrap and duct were coupled via lumping mass and surface j area as described in Reference 4.
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AL ISTA E FROM CORE BOTTOM (Z = 0.5") l Figure 2. XXO7 Instrumentation Loading Plan 8021-26 l 6 O l 1
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QQ@QQ QQ SUBAS BLY WALL THIMBLE FLOW LEGEND FM FLOWMETER LEAD CONDUlT 2 SPD SELF POWERED NEUTRON FLUX DETECTOR 2 BTC COOLANT TC CORE BOTTOM LEVEL 2 4TC COOLANTTC 0.137M ABOVE CORE BOTTOM LEVEL 2 7TC COOLANT TC 0.240M ABOVE CORE BOTTOM LEVEL 2 TTC COOLANT TC 0.322M ABOVE CORE BOTTOM LEVEL 9 ISTC COOLANT TC 0.514M ABOVE COR2 BOTTOM LEVEL 1 OTC COOLANT TC 0.933M AB0VE CORE BOTTOM LEVEL 2 FTC FUEL CENTERLINE TC 0.322M ABOVE CORE BOTTOM LEVEL 6 EL EXTENDED 0.6SM LENGTH MK -ll FUEL ELEMENTS 6 NOTES: 1. THE OUTLET COOLANTTCs(NOT SHOWN) ARE LOCATED NEAR THE SUBASSEMBLY COOLANT EXIT
- 2. THE 00TS REPRESENTS THE LATERAL POSITIONS OF THE TC JUNCTIONS IRRESPECTIVE OF AXlAL LOCATION Figure 4. Subawembly XXO8 Fuel Bundle Instrumentation Loading Plan 802i 8
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The model for XX08 assembly is similar, with the appropriate modifications due to the dimensional change, the difference in total element length and the presence of gas plenum in place of steel rod. The model for XX08 assem-bly is shown in Figure 6 The calculations for heat generation in the fuel element were based upon the assumption that the average element was entirely fuel material. Material properties such as conductivity, density, specific heat and gap coefficients were held constant and equal to those of fuel as listed in Table 4.1 of Appendix A for XX07 assembly and Table 4.2 of Appendix 8 for assembly XX08. Average fuel element power and mass velocity were input as functions of time. Thus, for XXO7 assembly, the nominal condition of 1.6% of design power and 5.35% of design flow were used as initial conditions. To account for r:.aasure-ment uncertainties, the power and flow were adjusted with 110% and 99% of the nominal conditions stated above. For the purpose of maintaining identical initial coolant temperature conditions and temperature rise across the core, the power and flow were either increased or decreased simultaneously to maintain a constant initial power / flow ratio. For XX08 assembly, calculations were made for the four elements near the assembly center. The mass velocity used the effective flow area of these channels and the average assembly flow rate divided by a flow maldistribution factor (FM9) which was obtained by the COBRA code (Ref. 5). Since the radial power distribution is relatively constant near center, the model uses the average power and flow of these four elements (23, 32, 40 and 31) in Figure 4. The axial power distribution, in discrete nodal representation, are shown in Figures 7 and 8 for assemblies XX07 and XX08, respectively.
- 4. FORE-2M ANALYSIS 4.1 XX07 Test F Analysis Using the initial power (7.2 kw) and flow (19.3 lb/hr) conditions with a constant core inlet temperature of 700 F, the FORE-2M calculation was made during the 120 seconds after the auxiliary pump tripped. The flow and power input transients are shown in Figures A-5 and 5.2 of Appendix A (Appendix A gives detailed information on input data and calculated coolant temperatures at core inlet, mid-core, core exit and assembly exit locations). The avail-able EBR-II Test F assembly XX07 measured data (Ref: 6) consists of plots of core inlet, mid-core, and core exit and assembly exit coolant tenperature transient.
Figure 9 shows the comparison of the FORE-2M calculated mid-core coolant tem-peature with the EBR-II data for the nominal power and flow case. The measured temperature peaked (865 F) at approximately 45 seconds into the tran-sient (or 24 min. 40 sec. after reactor scram). FORE-2M results indicated the maximum temperature of 870 F (5 F higher) occurred at about 7 seconds
- earlier. The calculated results also drop below the measured data after about 45 seconds into the transient and levels to about 20-25 F below the measured data. The ensuing discussion presents the reasons for this effect.
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1 1 l l 2.2 RELATIVE AX1AL POWER DISTRIBUTION FACTORS Fal2] 2.1 - F O R E.2M
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700 - 24 25 26 27 TIME (MIN.) Figure 9. Mid-Core Coolant Temperature Comparison (6.5" Abose Ilottom of Acti.e Fuel) FQRE-2M and EBR-II XX07 Test F O O O-
Figure 10 shows the comparison of the FORE-2M calculated core exit coolant (oV) temperature with the EBR-II data. Note two curves are given because the calculation yields sodium temperature at the middle of the top axial node (Z = 12.0 in.) and the exit of this node is at Z = 13.5 in. The top of core T/C on the other hand, was located at Z = 12.65 in. When this measured temperature is compared with calculated values, interpolation between the two dotted curves (FORE-2M results) should be used. The core exit temperature peaked at approximately 24 min. 52 sec. after the reactor scram, at 955 F. FORE-2fi results reach maximum about the same time, but overpredicts the peak temperature by 30"F. At eighty (80) secor.ds into the transient, the calculated temperature coincides with measured value and beccaes lower than the measured values after that time. Figure 11 shows the interpolated temperature pre-dictions of FORE-2M code at Z = 12.65 in. above core bottom and measured temperatur.95 at the same level. As reported in Reference 6, these are average of thirteen top-of-the-core coolant thern,ocouple readings. The range of individual measurements are shown by vertical bars over the data points. Considering the uncertainty of +10% in nominal power and flow conditions, FORE-2M calculations were made for 110% and 90% of these nominal values. The results cn top-ef-core temperatures are shown by dotted curves enveloping the nominal case. Also shown in Figure 11 is the predicted temperature from the NATCON code which is the primary system module of a whole plant simulation code NATDIMO (Ref. 7). NATCON models the reactor, outlet plenum and piping, IHX, pumps and the primary pool of EBR-II. The predictions from NATCON were based upon the perforuaace of an average driver (Ref. 2) and since the assembly flow of XX07 is less than average flow, it may be expected that measured temperatures from XX07 assembly were higher. The FORE-2M model overpredicts the maximum temperature by 30 F, or a maximum core temperature (m) rise of 285 F as compared to about 255 F in the EBR-II Test F data. After the C/ maximum top-of-core temperature peak occurs, the FORE-2M results drop below the test data. This is because the FORE-2M calculations of core temperatures used in this study do not incorporate transient flow redistribution and radial heat transfer effects within assemblies or between assemblies. These effects would tend to moderate the maximum temperatures and flatten the temperature pro fil e. Combined with the uncertainties in the power and flow measurements, this would explain differences between code predictions and the experimental data. Assembly exit coolant temperatures at Z = 37 in, were also presented in Figure A-10 of Appendix A. Poor agreement was shown with the measured exit tempera-ture. Several factors can be attributed to the poor agreement:
- 1) The thermocouple measurement location was at a point where significant inter-assembly mixing took place, which resulted in a lower and slower responding mixed outlet temperature;
- 2) The inter- or intra-assembly radial heat transfer is unaccounted for in the FORE-2M model used in this study; and
- 3) The analytical model essentially neglected the heat capacity effect in the steel rod region.
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24 25 26 27 TIME (MIN) Figure 10. Core Exit Coolant Temperature Comparison - EditE-23I and EBR-Il XXO7 Test F O O O.
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" F.6RE-2M. PREDICTION FOR XXO7 (NOMINAL CON 0lil0hS)
NATCON PRE 01CT10N FOR AVERAGE 800 FUEL ORIVER O AVERAGE OF 13 XXO7 TOP OF CORE
- COOL ANT TTC*S
- "" ~ ~ FBRE-2M PRE 0lCTIONS WITH +10%
UNCERTAINTY IN FLOW AND POWER INPUT 750 700 ! 24 25 26 TIME AFTER REACTOR SCRAE (MIN.) l I I I l l l l 0 30 60 90 120 TIME AFTER AUXlLIARY PUEP TRIP (SEC) Figure i1. Comparison of Top-of-Core Temperature in XXO7 Assembly During ) I Natural Circulation Test F 8021-3 17
4.2 XX08 Test 7A Analysis Observation of the data in Reference 5 which shows a rather constant flow mal-distribution factor (FMD) at rated steady state flows, and the fact that the radial power distribution in the assembly is relatively constant near the center lead to the assumption that the four "near center" elements and their measurements can be averaged. The FORE-2M central element, therefore, models the average power and flow of these four "near center" elements which are labeled in Figure 4 as 4TC-23, 7T-32,15TC-40 and TTC-31 and an axial distribu-tion of thermocouple locations is thus established. The FMD factor of 1.23 was used at the initial steady state condition. The COBRA-IV code (Ref. 5) was used to calculate the channel FMD factor under quasi-steady state conditions using Test 7A average power and measured flow as input (Figure B-8, Appendix B). The steady State axial temperature profile calculated by the COBRA-IV code and the FORE-2M calculated initial temperatures (using the average FMD factor) are compared in Figure 12. The EBR-II (XX08 assembly) Test 7A data measured near the assembly center are also indicated in this figure. FORE-2M and COBRA codes are go0d agreement (within 5 F) of each other, and are within 25 F of the measured data. Both coJes overpudict the maximum coolant temperature (or yield conservative predictions). In transier,t cases, the effect of flow distributica was simulated in the FORE-2M code via changes in flow input data by reducing the FMD. This FMD factor relates the !ccal enthalpy to average rod bundle enthalpy rise for a particular axial location. Four different transient cases of varying FMD factor for "near center" elements were calculated and the results are shown in Appendix B. Table 2 con-tains a brief description of each case calculated in Appendix B which includes alsc, cases of a variation in the element model (a near center element or a specific element) and reactor scram time. Of these cases, the last case of variable FMD (from COBRA-IV data) yields the most realistic calculational model. Figures 13,14 and 15 show the coraparison of the FORE-2M results with data. Figure 13 shows the coolant temperature at axial location Z = 5.5 inches. Also shown in this figure is the results of calculation for case },(a stepped change in FMD from 1.23 to 1.0 at flow rates less than 20% rated flow)\ 1 The variable FMD case at most is about 6 F above the test data or 7% of maximum temperature rise. Figure 14 is a comparison of coolant temperatures at 9.5 inches above the core bottom. The variable FMD case at most again is about 6 F above the test da ta . Figure 15 is a comparison of coolant temperatures at 12.68 inches above the core bottom (12.5 inches in FORE-2M calculation). The maximum deviation for the variable FMD case is about 15 F above the test data (or 14% of maximum temperature rise). In all these calculations, the results of case 3 calculations yields lower temperatures, which are below the test data most of the time. Figure 16 compares coolant temperatures at 20.5 inches above the core bottom. All calculated coolant temperatures at this location are higher than measured l ones primarily due to radial heat transfer outside of the effective fueled region. The above comparisons are made with "near center" elements. At the corners and flats, the difference between FORE-2M calculations and test data become larger for initial steady state conditions (Appendix B). Several factors may be contributed i to this: inter-assembly radial heat transfer and possible abnormal subchannel flow distributions. During the natural circulation transient conditions, near corner element measurements also snow larger deviation with calculated temperatures than that of near center elements. This is believed due to edge channel over-cooling. (*)The FitD = 1.0 condition would correspond to all coolant subchannels having an equal enthalpy rise for a particular axial location (i.e. , uniform radial temperature distribution across pin bundle). 18
O 940 900 - A TTC SSO - 4 C e, , [ 7TC E i- 810 - 5
$ J M
e- A
$ 780 -
4TC ' 5 o O 740 - COBR A-IV CALCUL ATIONS O (CHANNEL #3) A XX08 TEST DATA (NEAR ASSEMBLY CENTER) BTC,4TC,7TC,TTC AS SHOWN IN FIGURE 4. 700 - A O O FIRE 2M CALCULATIONS (1.23 FMD) 660 0 2 4 6 8 10 12 14 AXlAL -INCHES ABOVE CORE BOTTOM Figure 12. FORE-2M and Cobra Ceule Comparison of Asial Temperature 1)istributions With EllR-II Test 1)ata (7A) f 8021-4 O 19 i
- p. f
. e :
TABLE 2 FORE-2M TEST 7A NATURAL CONVECTION TRANSIENT CASES DESCRIPTION CASE NUMBER DESCRIPTION 1 Average channel. Pin power and flow average of sub-assembly total flow n;a1 distribution factor (FMD) = 1.0. 2 Near center of subassembly. Power and flow based on average of radia.1 power factor and average channel fion , maldistribution factor equal a constan_t. . FPD = 1.23; scram at 2.3 secords. i 3 Saw as Case 2 except that flow maldistribution factor step changed to one (1) at flow rates less than 26% cf design f'cw. 1.0 < FMD < l.23; scram at 2.3 reconds. 4 Sane as Case 3 except reactor scram occurs at 1.8 l seconds after pump trip. : 5 Flow, FMD factor,' and power calculated as per pin position 10. TTC = 5.4 in, above core bottom. j 6 Flow, FMD factor and power calculated as per pin position 17. TTC = 9.5 in. above core bottom. 7 Flow FMD factor and power calculated as per pin position 16. TTC = 12.68 in. above core bottom. 8 Same as Case 2 except FMD = F(T) per Figure B-8 of -, Appendix B. t t 1
! O O -
i i 8 no I f A FMD STEPPED FROM 1.23 TO 1.0 AT 10 SEC. O F9RE-2M FMD VARIABLE TEST 7A DATA d 800 1 NOTE: CASE 8 (SEE TABLE 2) 1 ) o[ 760 - m E
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=
s ! S g 720 - i 4 -g g .! a a OO 680 - l l I I I 640 0 20 40 60 80 100 120 TIME (SEC) Figure 13. F9RE-2M With FMD and EBR-II tXX08) Test 7A Coolant Temperature Comparison at Axial Location Z = 5.5 inches I
;.d
4 9
960 A FMD STEPPED FROM 1.23 TO 1.0 AT 10 SEC 0 920 - O F9RE-2M FMD VARIABLE TEST 7A DATA NOTE: CASE 8 (SEE TABLE 2) 880 - A 840 -- C o_ IU
=
Q 800 - e: l E 3 O 760 - O N 720 - O A 680 - 640 0 20 40 60 80 100 120 TIME (SEC) Figure 14. FORE-2M With FMD and EllR-II (XX08) Test 7A Coolant Temperature Comparison at Asial Location Z = 9.5 inches 8021-23 22 0
. 960 0
920
-a A FMD STEPPED FROM 1.23 TO 1.0 AT 10 SEC j,
O FDRE 2M FMD VARIABLE TEST 7A DATA 880 - NOTE: CASE 8 (SEE TABLE 2) 840 -- C
~
w 8 Q 800 - 5 NO 760 - ( O O g 720 o ao 680 - 640 ! ! 0 20 40 60 80 100 120 TIME (SEC) i Figure 15. FORE-2M With FMD and EBR-II (XX08) Test 7A Coolant Temperature Comparison at Axial Location 7. = 12.68 inches 8021-24 O 23
960 a 920 - A FMD STEPPED FROM 1.23 TO 1.0 AT 10 3EC O FBRE 2M FMD VARIABLE TEST 7A DATA 880 - . NOTE: CASE 8 (SEE TABLE 2) 840 - C c E e t 4 800 -, 5 03
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a O 760 - o a o a a N l %~~ O l A I 720 - 0O l O/ 680 - 640 0 20 40 60 80 100 120 TIME (SEC) Figure 16. EpRE-2M With FMD and EllR-il(XX08) Test 7A Coolant Temperature Comparison at Asial Location Z = 20.5 inches l 8021-22 1 24 0
To show the effects of inter-assembly heat transfer and intra-assembly flow and heat redistribution, COBRA-WC code predictions are presented (Ref. 3). Figure 17 shows the comparison of XX08 test data with the FORE-2M model and the COBRA-WC model. The FORE-2M model uses the flow, FMD factor and power calculated as per element 17 (case 6 in Table 2). The initial steady state temperature is about 25*F lower than the measured value while the COBRA-WC calculated value is about 50 F lower. During the transient, the FORE-2M calculated temperature is 25'F (or 25% of the maximum temperature rise) higher than the measured values at the peak temperature. The COBRA-WC model, on the other hand, results in good agreement with the data by considering the inter-assembly heat transfer and intra-assembly flow and heat redistribution. Figure 18 shows the similar comparison with element 32 which is near center. The steady state temperature is very close to the measured and the largest temperature difference of about 15*F (or 15% of maximum temperature rise) occurs at the time of maximum tem-perature. Again, without considering the inter-assembly heat inn fer and , intrs-assembly flow and heat redistribution, the FORF-2H calculatica cver-predicts toe coolant transient temperature. With these factors considered in the COBRA-WC model, the temperature predictions are reduced and closer to the ' measured values.
- 5. CONCt.USIONS Selected EBR-II experimental data on assemblies XX07, XX08 were cotpared with ,
FORE-EM code calculations. Assembly averaged values are used for XXO7 compari-son. As such, the heat, pressure drop and flow redistribution within the assembly and the inter-assembly heat transfer were not considered. In the case b V of XX08 assembly data, near center subchannels are treat 2d, the intra-assembly flow redistribution was simulated by varying the FHD factor. In general, the FORE-2M model over-predicts the maximum core coolant temperature reached during the natural circulation transients. By considering the flow redistribution alone, the FORE-2M predictions agree very well with XX08 assembly test 7A data in "near center" subchannels. The largest differences in coolant temperature comparison cases occur when near corner element measurements were compared with FORE-2M results. Edge channel overcooling by inter-assembly heat transfer and intra-assembly flow redistribution are the causes of the deviation. When these factors are taken into account, as was done in the COBRA-WC model, the agree-ment was improved. A FORE-2M capability has been developed for natural circulation calculations to incorporate these factors (Ref. 8) ir this level of detail is desired by the user. Considering the power and flow measurement uncertainties of about +10% in these tests results in a band of 10-20 F variation in the measured coolant tem-perature. Therefore, the FORE-2M results are within the region of test data uncertainty. More importantly, in all cases FORE-2M predictions of maximum temperatures are conservative.
- 6. REFERENCES
- 1) R. D. Coffield and H. P. Planchon, "LMFBR Natural Circulation Verification Program (NCVP) Review of Experimental Facilities and Testing Recommendations",
WARD-NC-3045-1, July 1977. 25
8 9 A ale D b N I I; y I I a I 8
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- 2) J. L. Gillette, J. E. Sullivan, R. M. Singer, et al, " Compilation of Data from EBR-II Natural Circulation Test Procedure EX-140, Section IV, Addendum 8", ANL/EBR-112, May 1980
- 3) E. U. Khan, W. A. Prather, T. L. Georca, et al, "A Valida' tion Study of the COBRA-WC Computer Program for LMFBR Core Thermal-Hydraulic Analysis", Part I, COBRA-WC Generalized Parameters for Steady-State Analysis, PNL-4128, December 1981
- 4) J. V. Miller and R. D. Coffield, " FORE-2M: A Modified Version of the FORE-II Computer Progra:n for' the Analysis of LMFBR Transi-ents", CRBRP-ARD-0142, Novembe r 1976 -
- 5) C. L. Wheeler, C. W. Stewart, R. J. Cena, et al, " COBRA-IV-I Interim Version of Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores", BNWL-1962, March 1976
- 6) J. L. Gillette, D. Mohr, R. Singer, and R. Smith, "A Flow Coast :
down to Natural Convection Conditions in EBR-II", Trans. Am. Nucl. Soc. 22, p. 594, (1975) ,
- 7) D. Mohr and E. E. Feldman, "A Dynamic Simulation of the EBR-II Plant _ -
during Natural Circulation with the NATOEMO Code", Decav Heat Removal and Natural Convection in FBR's, p. 207, Hemisphere Publishing Corp., , 1981
- 8) R. D. Coffield, J. S. Killimayer, Y. S. Tang, and R. A. Markley,
" Buoyancy-Induced Flow and Heat Redistribution during LMFBR Core De-cay Heat Removal", Ibid, p.177
- 9) J. V. Miller, R. D. Coffield, K. D. Daschke, et al, " Supplementary Manual for the FBRE-2M Co ;puter Program", CRBRP-ARD-0257, tiay 1982 l
l l 1 l i i 1 28 O
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I I ? I 4 4 l 1 I 4 I I l 4 i i i ) i i I ' s > l , i i ? l l APPENDIX A
) '
f i COMPARISON OF F9RE-2M NATURAL CIRCULATION TRANSIEfiT PREDICTIONS tlITH EBR-II, , !, XX07 SUBASSEE LY, TEST F DATA ! l 4 i s . I l 1 i' 1 1 l l $ t , I l ' l 4 4 r 4 l 1 1 J i 6
I 1 1 .I i t, l TABLE OF CONTENTS , i Pigg, i l
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . A-1 I 2.0 EBR-II TEST F DESCRIPTION . . . . . . . . . . . . . . . . . . . A-1 3.0 X X 0 7 S U8AS S EMB LY DES I GN . . . . . . . . . . . . . . . . . . . . A- 2 i I 4.0 F9RE-2M ANALYTICAL MODEL OF AVERAGE CHANNEL . . . . . . . . . . A-7 . r ' l i 5.0 FPRE-2M TRANSIENT ESULTS . . . . . . . . . . . . . . . . . . . A-12 . i
6.0 CONCLUSION
S . . . . . . . . . . . . . . . . . . . . . . . . . . A-26 i J
7.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . A-27 ' i 4 L l E j J l i , i i 1 l l l I i A-1 i
LIST OF FIGURES Page A-1 XX07 INSTRUMENTATION LOADING PLAN. . . . . . . . . . . . . . . . A-3 A-2 EBR-II ENVIRONMENTAL INSTRUMENTED SUBASSEMBLY XX07 . . . . . . . A-4 A-3 SCHEMATIC 0F FORE-2M AVERAGE CHANNEL N0DAL MODEL . . . . . . . . A-8 A-4 XX07 AND FORE-2M AXIAL POWER DISTRIBUTION. . . . . . . . . . . . A-9 A-5 XXO7 TRANSIENT FLOW BEHAVIOR - TEST F. . . . . . . . . . . . . . A-14 A-6 XX07 TRANSIENT POWER - TEST F. . . . . . . . . . . . . . . . . . A-15 A-7 CORE INLET COOLANT TEWERATURE COMPARIS0N WITH EBR-II TEST F DATA . . . . . . . . . . . . . . . . . . . . A-16 A-8 MID-CORE COOLANT TEMPERATURE C0WARIS0N WITH EBR-II TEST F DATA ....................A-18 A-9 CORE EXIT COOLANT TEWERATURE COMPARIS0N WITH EBR-II TEST F DATA . . . . . . . . . . . . . . . . . . . . A-19 A-10 SUBASSEf0LY EXIT COOLANT TEW ERATURE C0WARIS0N WITH EBR-II TEST F DATA ............. . . . . . . . A-20 A-ll CORE INLET COOLANT TEMPERATURE WITH +10% UNCERTAINTY IN POWER AND FLOW COMPARED TO EBR-II TEST F DATA. . . . . . . . . . . . . A-22 A-12 MID-CORE COOLANT TEMPERATURES WITH +10% UNCERTAINTY IN POWER AND FLOW COMPARED TO EBR-II TEST F FATA. . . . . . . . . . . . . A-23 A-13 CORE EXIT COOLANT TEWERATURES WITH +10% UNCERTAINTY IN POWER l AND FLOW COMPARED TO EBR-II TEST F DATA. . . . . . . . . . . . . A-24 A-14 SUBASSEMBLY EXIT COOLANT TEMPERATURES WITH +10% UNCERTAINTY IN POWER AND FLOW COMPARED TO EBR-II TEST F DATA . . . . . . . . A-25 l O l A-ii l 1
i J j - i LIST OF TABLES .i 1 I ?agt j i A-1 XXO7 SUBASSEMBLY CHARACTERISTICS. . . . . . . . . . . . . . . . A-5 A- 2 XXO 7 P ROPE RTY TAB L E . . . . . . . . . . . . . . . . . . . A-10 A-3 F9RE-2M AVERAGE CHANNEL INITIAL CONDITIONS. . . . . . . . . . . A-13 i i i j t ! l i I l l I i i 1 ( l I t l l l A-lii I ( ..-- - .. - , - - -.-- ,.. ..._,- - - ._ _ _. . . . . - - - . . _ . . , _ . . _ _ . _ . . . . - - - - _ - _ . . . - - . . _ - - - _ . - . _ - - ,
O
1.0 INTRODUCTION
This report presents comparisons of F9RE-2M computer code [A-1]results with EBR-II natural circulation test data. Also, a brief description of the EBR-II natural circulation test, the analytical model, assumptions and input data used to simulate the EBR-II XX07 subassembly themal response during natural circulation conditions are presented. 2.0 EBR-II TEST F DESCRIPTION On September 24, 1974 a natural convection experiment, designated as XXO7 Test F, was conducted in EBR-II. Test F was initiated after an anticipatory O shutdown scheduled for the end of reactor nm 73 which had completed 2099 mwd operation (44 days). The reactor was shut down by inserting the highest worth control rod at its nomal drive speed (5 in./ minute). When this rod reached its zero bank position, the core was then subcritical and the remain-ing control rods were scrammed. The primary coolant pumps were maintained at full flow for 5 minutes after the reactor was scrammed before they, in turn, were tripped. The resultant flow dropped to approximately 5% (of the rated flow) due to the continued operation of the auxiliary pump. The aux-iliary pump flow was maintained for 15 minutes after the primary pumps were tripped, and following the auxilicry flow coastdown, all primary system flow was due to natural convection. Approximately 120 seconds after the auxiliary pump trip, Test F was teminated by restarting the primary pumps. O A-1
3.0 THE XX07 SUBASSEELY DESIGN The XX07 design is that of an instrumented fueled subassembly located in a converted control rod position. A cross-section of XX07 is shown in Figure A-1 and illustrates the location and types of instrumentation employed. The entire subassembly configuration and shape is shown in Figure A-2. The subassenbly instrumentation included two inlet flow meters, ten fuel center-line thermocouples, three inlet and two outlet coolant thennocouples, three core midplane coolant thermocouples, and five coolant thermocouples near the top of core. The five latter thermocouples were placed approximately 0.85 in, down from the top of the feel position so that they would be at the same axial level as the fuel centerline thermocouples. All of the elements in the subassembly were fueled except for four: two elements were used as conduits for the flow meter leads which passed through the fueled region of the subassembly, one contained both a rhodium and a platinum self-powered detector (SPD) and one contained thermal-expansion difference tem-perature monitors (TEDS). The XX07 subassembly contains 61 elements each with an outer diameter of 0.174 in, which were spiral wrapped with 0.049 in. spacer wire utilizing a 6.0 in, helical pitch. The fuel has a 0.144 in. outer diameter,13.5 in. long pins which are sodium bonded to the stainless steel cladding. The clad-ding and sodium bond thicknesses are 0.009 in. and 0.006 in., respectively. The interior dimension of the inner hex can is 1.83 in. across flats and the hex can is 0.040 in. thick. The subassembly sets in an INSAT ) facility I 1 which, in the core region, consists of a hexagonal guide tube which is also l 0.040 in. thick. The gap, a thimble region between the guide tube and XX07 hex can is 0.15 in. A sumary of these dimensions is given in Table A-1. The effective length of the fuel elements is 36.0 in. This length is com- ! prised of a 0.5 inch coolant entry region (containing a stainless steel reflector and grid support attachment), a 13.5 inch fueled region, 4.0 inch gas plenum space and an 18.0 inch stainless steel upper reflector / shielding region. The wire wrap is attached at the beginning of the fuel element
)An instrumented subassembly that provides continuous in-core monitoring.
O A-2
TO C9RE CENTER GRID BAR ORIENTATION e 900
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"S' "
li'eU?f!"nfSU!"' ' - TEMP MONITORS
- TED 1 -
AL ISTA CE FROM CORE BOTTOM (Z = 0.5") Figure A-1. XX07 Instrumentation Loading Plan 8021-26 O A-3
THE WIRE WR AP STARTS AT THE 215 MARK AT THE START OF THE ELEMENT (Z = 0) SUPPORT
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FCTC 51 FUEL ELEMENT ? g ARRANGEMENT (2X) SHOWING FUEL CENTERLINE $ dp[ END FITTIN_Gh n PPER THE RMOCOUPLES (FCTC) > THE WIRE WRAP 9 SOT SUBASSEMBLY Oo DIRECTION IS OUTLET COUNTERCLOCKWISE THERM 0 COUPLE 36" 2858 (Z=0) _ 1/4, 908 270 H i UPPER
.,_ '3 A D APTER -+
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s SUPPORT-i (((((([j[ u l ~"y g --; ge GRID n c y y y t g3g-
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" 5 COOLANT ::;':: l 0625 ) FUEL ELE?AENT Figure A-2. EllR-ll EnvironmentalInstrumented Subassembly XX07 8021-28 A-4
O TABLE A.1 XXO7 SUBASSEMLY CHARACTERISTICS SUBASSEMBLY PARAMETERS DIMENSION Hex Can, Flat-to-Flat ID 1.83 in. Hex Can Wall 0.04 in. Thimble Gap (Between Hex Walls) 0.15 in. Pin Diameter, OD 0.174 in. Wire Wrap OD 0.049 in. Wire Wrap Pitch 6.0 in. Sodium Bond Thickness 0.006 in. Cladding Thickness 0.009 in. Pin Number 61 Wire Wrap Zero Reference Angle (Start of Fuel, Z = 0.5 in.) 315* STEADY STATE FULL FLOW CONDITIONS VALUE Total Bundle Flow, lb/hr 22.06 x 10 3 Total Thimble Flow, lb/hr 2.30 x 10 3 Average Element Heat Flux (Over 36 in. of Length) Btu /hr-ft2 t 1.8 x 10 5 Inlet Temperature, *F 700 Inner Hex Cgg verag: Heat Flux, l Btu /hr-ft i 2.8 x 10 3 1 Total Subassembly Power, kw 439.7 Total S/A Pressure Drop (Inlet and Outlet Orifices Included), psi 37.4 N It was assumed that this heat flux was equall inner (bundle side) and outer (thimble side) yhex divided over both can faces. O A-5 O
(Z = 0) in the 285 position (i .e. , the 0 position is North, the 90* post-tion is West, etc.) and spirals in a counterclockwise direction. The wire wrap rotates 60 per inch and at the start-of-core positions is at the 315 mark (Z = 0.5 in.). See Figure A-2 for additional details. The steady state full flow conditions in XX07 prior to Test F consisted of a 700 F inlet temperature with a total bundle flow of 22,060 lb/hr. Note that af ter passing through the last flow meter, the coolant experiences a flow expansion region before entering the bundle as illustrated in Figure A-2.. The coolant also experiences another expansion region at the end of the 36 in. element length before being mixed further by the upper reflector assembly in the upper adapter area. O 1 l I A-6
4.0 FORE-2M ANALYTICAL MODEL OF AVERAGE CHANNEL , The FORE-2M code solves steady state and transient thermal-hydraulic reactor core performance. Since the information available at this time is essentially EBR-II subassembly parameters and average subassembly temperatures, the analytical model was , chosen to represent an average pin / channel in the XXO7 subassembly for com-parison of FORE-2M with EBR-II XXC7 subassently test data. The key core i parameter data used in g:;nerating the model is listed in Table A-1. The XX07 fuel pin consists of 13.5 in. of fuel in the active core region and contains a steel rod of approximately 22.0 inches extending above the fuel to the subassembly exit region. The FORE-2M fuel pin model is divided into seven axial and 5 radial nodal sections as illustrated in Figure A-3. Six axial nodes are used to model the active core (fuel) region and one axial node is used to represent the steel rod including the coolant entry region. The radial nodalization con-sists of three equal volume fuel nodes, one cladding node with sodium bonded gap and a coolant node. The wire wrap and duct were coupled via lumping mass and surface area as described in Reference [A-1]. Initially, the radial fuel model consisted of six nodes but was reduced to three nodes when transient result comparisons, using nominal conditions as defined in section 5, showed less than 0.1 F difference in transient temperature reductions. The axial power distribution data was taken from data given in Reference A-3, and shown in Figure A-4 Also shown is the FORE-2M power distribution approximations. As shown, the FORE-2M data has constant power per each axial node region. The calculations for heat generation in the fuel pin were based upon the assumption that the average pin was entirely fuel material. Therefore, properties such as conductivity, density, specific heat and gap coefficierits are held constant and equal to those of the XXO7 fuel as listed in XX07 properties (Table A-2 ). To conservatively account for the steel rod portion of A-7
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i 4 i 6 i 8 i 10 i 12 i 14 s 36 AXlAL DISTANCE FROM THE BOTTOM OF THE PIN BUNOLE (IN.) l 1 i Figure A-4. XXO7 and F$RE-2M Axial Power Distribution l l 8021-14 I O A-9
TABLE A-2 O XX07 PROPERTY TABLE PROPERE ACTIVE CORE FUEL Thermal Conductivity, Btu /hr-f t *F 20.0 Specific Heat, Btu /lb *F 0.0445 Density, lb/f t 3 1107.0 CLADDING Thermal Conductivity, Btu /hr-ft- F 13.0 Specific Heat, Btu /lb *F 0.14 Density, lb/f t 3 481.0 Cladding Thickness, in. 0.009 BOND Bond Heat Tgansfer Coefficient, Btu /hr-ft F 9999.0 0 A-10
the pin, 4 the sodium bond heat transfer coefficient was reduced by a factor of 10 , (i.e., the gap coefficient set to zero), and the heat generation set to near zero. The sodium coolant properties varied as functions of tempera-tures according to property tables contained in the F9RE-2M computer code . Pin geometry and physical dimensions such as length, thickness and gap sizes remained constant throughout the thermal transients. Total pin power and mass velocity were input as functions of time. Therefore, heat, pressure drop and flow redistribution calculations are not considered. The effective flow area for the mass velocity is based on the average XX07 subassembly void flow area and average XXO7 subassembly flow rate. The nominal conditions of 1.6% of design power and 5.35% of design flow were used to establish the FORE-2M model initial conditions for the natural circulation transient calculations. To account for measurement uncertainties and to determine thermal sensitivity, the power and flow were adjusted +10% _ of the nominal conditions described above. For the purpose of maintaining identical initial coolant temperature conditions and core AT, the power and flow were either increased or decreased simultaneously to maintain a constant initial power / flow ratio. O A-ll
5.0 F,p,RE-2M TRANSIENT RESULTS The initial conditions for the natural circulation transient comparison analy;is with. EBR-II Test F data is summarized in Table A ! The initial power is 1.6% of full power (7.2 kw) and the flow rate is 5.35% of full flow (19.3 lb/hr). These conditions yield a core temperature rise of 55'F, and a 56 F temperature rise across the entire subassembly. l ( The core inlet temperature is 700*F and is assumed constant throughout the natural circulation transient. The flow and power input transients are shown in Figures A-5 and A-6, res-( pectively. As shown in Figure A-5, the decay powe changes very little over the two minute natural circulation test period (less than 4% of the initial power level). As shown in Figure A-6 the flow transient is similar to a step change with
, an initial change of flew to s0.7% of total flow (which cormsponds to a relative change of 85% of initial flow) and recovers due to buoyancy effects to about 1.35% of total flow. (Note: the scatter points in the figure are due to noise in instrumentation - Reference A-3).
The initial conditions and puser and flow transients described above are the nominal conditions for the FORE-2M computer code natural circulation trantient comparison case. The available EBR-II Test F subassembly XX07 measurement data [A-23 consists of plots of core inlet, mid-core, core exit and subassembly exit coolant temperature transient. Figures A-7 throughA-10 show the compari;on of the FORE-2M computer code temperatures with the EBR-II data for the nominal power ar.d flow case. Figure A-7shows the core inlet temperature corresponding to a measurement location of 0.75 in, above the core bottom. The temperature increases at a higher rate than the test data, and overpredicts the maximum teinperature by A-12
r i . . l ..
- O i TABLE A-3 F#RE-2M AVERAGE CHANNEL INITIAL CONDITIONS POWER, kw 7.2
! FLOW, lb/hr 19.31 COOLANT TEMPERATURES, *F: 1, CORE INLET 700 i MID-CORE 727 CORE EXIT 755 i i ASSEPBLY EXIT 756 i lO l 1 (*)See Figure A-3 for schematic of average channel. i 4 i l A-13 ) 1
. ._ --- -_.._-_-____..__._-..--__-- - - - - _ ._-- .- -l
1.01 NOTE: 1.00 NORMAll2ED POWER FOR F9RE 2M
$9s INPUT POWER TABLE e
No Ne 0.99 N,Ne a:
$ 0.98 -
E S e Y
! 0.97 -
e 0.96 - e TEST DATA EBR il TEST F 0.95 - 0.94 0 25 50 75 100 125 150 l TIME (SEC) l l l Figure A-5. XXO7 Transient Power for Test Fl A-3l 8021-21 O, 1 A-14 1 l l l
l 1 x 6 - o tJ ea
. .. . . .. XXO7 UPPER FLOW METER - CH.(74) c *
....g.,,.-..*..
- ris ., EBR4I
, if . .
XXO7 TEST F
.,.s..*,....,...a v..
o e <.--
... ,,'o -
5 - 4 - Z uJ L3 KC _E 3 i
> =
a L a 2 -
..f . /.._ . .
.' uv.~ :.
_i ^ [l<._....s_
. . . . s....,
, I I I I I I I I 22.0 22.5 23.0 23.5 24.0 24.5 25.0 25.5 26.0 25.5 TIEE (EIN.)
Figure A-6. XXO7 Transient Flow Behavior for Test Fl A-2l
E L" de 800 BTC 18-CH(78) F8RE 2M C EBRll
- MEASURED g ,, _ _
AVERAGE
. =.
o
$ 2 W
(*IINLET DEFINED AS EXIT DF FIRST NDDE 600 I I 24 25 26 27 TIME (MIN. ) 1 Figure A-7. Core inlet
- Coolant Temperature Comparison - EBR-II XXO7 Test F O O O-
r approximately 8 F. At about 40 seconds into the transient, the temperature levels to s5*F below the test data. Figure A-D shows the comparison of mid-core temperatures which correspond to a measurement location of 6.5 in. above the bottom of active fuel. As shown, the maximum temperature occurs about 7 seconds sooner and about 5*F higher in the FORE-2M results. The F9RE-2M results also drop below the measured data at about 50 seconds into the transient and levels to about 25 F below the EBR-II data. Figure A-9 shows the core exit coolant temperature calculated at 13.5 in. (Node 6) above the core bottom. However, the actual temperature measure-ment is located at 12.6 in. above the core bottom. Therefore, the FORE-2M average (Node 6) temperature located at 12.0 in. above the core bottom is also shown in Figure A-9 Hence, the actual FORE-2M results lies about midway between the two data curves plotted. As shown, the FORE-2M results reach maximum temperature a few seconds earlier than the test data and over-predicts the maximum temperature by 20*F and 40*F for the nodal average and core exit coolant temperatures, respectively. Ninety (90) seconds into the transient, the temperatures drop to 0*F and 20*F below the measured data. The interpolation of the 12.0 in. and 13.5 in. FORE 2-M temperatures (i.e.12.75 in. location) yields a maximum core AT of about 280*F as compared to about 255*F in the EBR-II Test F data. Figure A-10 shows the FORE-2M subassembly exit temperature compared to EBR-II test data. The FORE-II transient initially drops 30*F (maximum) below the test data. However, 30 seconds into the transient, it starts increasing at an average rate of 5'F per second. This rate corresponds to the same rate as the active core coolant temperature rate (Figure A-9). As shown in Figure A-10 the maximum temperature occurs 30 seconds earlier and about 40 F higher than the test data. Figures A-ll through A-14 show FORE-?M core coolant tempera'ture comparisons with EBR-II test data for the two cases consisting of changing power and ficw +10% of nominal Test F Data. Each figure contains a vertical bar which A-17
8 1000 _r. 900 - ! C l o. F5RE-2M
'~
E /
? s 4 / N f
's s (xxo7 ME ASURE0 Z = 6.5")
> w w s b " /
* ' ' ~ ~ ~ ~ - FWRE-2M AT 6.5" ABOVE BOTTOM 800 -
/ OF ACTIVE FUEL
/
/
/
/
/
70fl 24 25 26 27 TIME (MIN.) Figure A-8. Mid Core Coolan! Temperature Comparison (6.5" Above Bottom of Active Fuel FWRE-2M and EBL-II XXO7 Test F O O O-
O O O- . 8 1000 ,% Lf EJ TTC 20-CH(80) { FWRE-2M (13.5")
//,n \ .
\ ME ASURE D - 12.55"
/ g .
,/ g ABOVE CORE BOTTOM
/ \
N 900 's
~~~_
/
g c- / F NRE-2M (12.0") E . ? n. PEAK 3TEBR-il = 255'F 2> g PEAK JTFf RE-2M " 388'E
# f G
800 -
----NODE 5 AVERAGE TERIPERATUK f
/ .-. -NCDE E EXIT TERAPERATURE i I I I 24 25 26 27 TIME (MIN.)
Figure A-9. Core Exit Coolant Temperature Comparison - FWRE-2M and EBR-Il XXO7 Test F
g 1000
' d. MEASURED ASSY XXO7 OTC 19 CH(82) h FERE 2M PREDICTION NATCON PREDigDN FOR AVERALE CRIVER ASSY W OF CORE AT -
f,N* $
\ F B R E-2M 900 -
7 , ME ASURE D E BR11
/ (
= . i i
g -
/ i l
> 3 l 6 [ i 800 -
/ I
./ I
/ s
%. / '
I 700 i l I\- ' 24 25 26 27 TIME (MIN) Figure A-10. Subassembly Exit Coolant Temperature Comparison - FORE-2M and EBR-il XXO7 Test F O O O-
O represents.the affect on temperature due to the 110% (uncertainty) change in measured data. The upper portion of the bar represents the 90% of nominal power and flow adjusted input data, and similarly the bottom portion of the bar represents the 110% of nominal power and flow adjusted input data. Figure A.111s a comparison of the FORE-2M and EBR-II core inlet temperature. and shows only a 11?F maximisn variation in calculated temperature for a 110% variation in power and flow. Figure A-12 is a comparison of the F9RE-2M and EBR-II mid-core temperature and shows a maximum variation of 17'F. Figure A-131s a comparison of F9RE-2M and EBR-II core exit temperature and shows a maximum variation of about il3*F. Figure A-14 is a comparison of F9RE-2M and EBR-II subassembly exit temperature and shows a maximum temperature deviation of u_20'F about the nominal case. In all cases, the maximum deviations occur just prior to reaching maximum. coolant temperatures, and diminish to a minimum (<5*F) after the temperature peak occurs. O A-21 1
tJ 800 BTC 18-CHl78) EBR11 MEASURED
~_________________q g
E 70g - (
> 8 L
to
~
NOTE: I + 1 F8 VARIATION FOR 90% AND 110% ADJUSTMENT ON POWER AND FLOW 600 24 25 26 27 TIME ( MIN.) Figure A-II. Core Inlet Coolant Temperature Comparison With +10'7r Uncertainty in Power and Flow - FORE-2M and EBR-il XXO7 Test F l , I \ O O O.
_ _ ..-. =_ .. _ , _ _ _ - _ _ . - _ _ _ _ .__- ___ b o 0- . !
,E
_s I 900 8 MAXIMUM DEVIATION +7 F AT 30 SEC. FOR 90% ' AND 110% ADJUSTMENTIN POWER AND FLOW l~
/- Nr
& s
- / -r
's EBR11 E ) 'I~
- s I Q 800 -- / % ' * ' 'I -I- - FORE 2M E
k T* Y O l 7,, I I I 24 25 26 27 TIME (MIN.) Figure A-12. Middle Active Fuel Length Coolant Temperature Comparison With +10'd Uncertainty in Power and Flow - FORE-2M and EBR-II XXO7 Test F
x 1000 i3 TTC 20-CH(80)- 12.65" ABOVE CORE BOTTOM r
\
P
/ N
,[ N - E BR-il (Z = 12.65 INCHES) 900 -
/ \
o' FORE 2M (Z = 12.0 INCHES) l-
~
N E E e s i' Y / Z 800 - /
/ -
/
\
700 24 25 ' 26 27 TIME ( MIN.) Figure A-13. Core Coolant Temperature Comparison With + 10'1 Uncertainty in Power and Flow qZ = 12.65 In.)- FORE-2M and EBR II XXO7 Test F , 4 O O O.
i
$ 1000 U j
,j OTC 19 CH(82) 4
% FDRE-2M 900 -
oI . EBR11 l l 5 I f l
< g 5
t f' I
? W / '
i m / I 800 -
).'
. .l A.
m /
%}d 700 ! !
24 25 26 27 TIME (MIN.) Figure A-14. Subassembly Exit Coolant Temperature Comparison With +10'; Uncertainty in Power and Flow - FORE-2M and EBR-II XXO7 Test F
l l .
6.0 CONCLUSION
S In general, the FORE-2M model over-predicts the maximum core coolant tempera-ture achieved during the natural circulation transients. However, after the l maximum temperature peak occurs, the FORE-2M results drop below the test data. This is because the FORE-2M calculations of core temperatures include heat, capacity of cladding, fuel, wire wrap and inner duct, but do not incorporate transient flow redistribution and radial heat transfer effects within assem ' blies or between assemblies. These effects would tend to moderate the maximum temperatures and flatten the temperature profile as indicated by the 10% . adjustment on flow shown in Figures A-ll through A-13. This is illustrated by observing the lower point of the vertlical bar, which gives a lower temperature profile as well as a Tower maximum temperature. Also, after the peak tempera-ture occurs, the +10% power and flow ! adjustment has less effect on the temperatures predicted by FORE-2M. ihe maximum temperature difference occurs at the core exit and is 40 F at maximum temperature, which corresponds to
. a maximum core AT of $300'F, +7 F (+10% uncertainty) compared to EBR-II Test . F data.[A-2] '
The rather poor comparison of the subassembly exit coolant temperature can be O attributed to several factors: firsti this region is modeled using fuel properties rather than stainless steel properties. Next, the heat capacity effect is essentially neglected in FORE-2M (i.e., zero gap conductance) in i the steel rod region (node 7 only). .The added heat capacity would delay the , FORE-2M temperature rise. Also, the thermocouple measurement location is j where significant inter-assembly mixing takes place yielding the lower and
' slower responding mixed mean outlet temperature. These coolant mixing effects' are, also, not modeled in the FORE-2M code. Finally,andmostimportantly,is[
the radial heat transfer either inter- or intra-assembly is unaccounted for ' in the FORE-2M channel model. All these factors could add delay to the FORE-2M temperature response at the subassembly exit a..d would yield a better thennal transient comparison in this region. O A-26
t I i i 7.0 R..E__FERENCES A.I. J. V. Miller, R. D. Coffield, "FSRE-2M: A Modified Version of the i F#RE-II Computer Program for the Analysis of LMFBR Transients", { CRBRP-ARD-0142, 1976. I
! A-2. J. L. Gillette, D. Mohr, R. Singer, R. Smith, "A Flow Co3stdown to i Natural Convective Conditions in EBR-II", ANL-IDAH0, Trans. Amer. Nucl .
! Soc. , 22, p. 594,1975. A-3. Personal Communication with P. R. Betten of ANL,1/29/79. i i i 4 e i 4 a i e 4 i I r i i i } A-27 4 1 c__,.__. . _ _ _ _ _ _ - - _ - _ _ . _ . _ _ _ _ . . . _ - . _ .-
l l. i s I 4 1 t .1 a e 1 t i i i. i ,' APPENDIX B f a l
; COMPARISON OF FORE-2M COMPUTER CODE NATURAL CONVECTION i ! RESULTS WITH EBR-II, XXO8 SUBASSEMBLY, TEST 7A DATA ;l l
c j l 5 i
!O 1
i i I
~ _ - . - _ - - - - - - . . _ _
-.&*.-A.- .a - .h-.. a-_m. ---& L -.e *ah-- .6% ..a h e- L- --.a-.. JAAJ #a.4a.. 2- - a a ,, aa .
TABLE OF CONTENTS 't h
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . 3-1 1 2.0 EBR-II TEST 7A DESCRIPTION. . . . . . . . . . . . . . . . . . . B-1 . J 3.0 XX08 SUBASSEMBLY DESIGN . . . . . . . . . . . . . . . . . . . . B-2 4.0 XX08 TEST 7A DATA . . . . . . . . . . . . . . . . . . . . . . . B-7 5.0 FORE-2M ANALYTICAL MODEL OF PIN / CHANNEL . . . . . . . . . . . . B-12 1 i 6.0 FORE-2M TRANSIENT RESULTS . . . . . . . . . . . . . . . . . . . B-19
7.0 CONCLUSION
S . . . . . . . . . . . . . . . . . . . . . . . . . . B-38 i
8.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . B-40 a I. i I 8 i
}
i i I ] . B-1 4 I i l
,-- -, ..., , . , -. ,, , . - - . . - . ,. .... . , . , . - . - . . ... ~ . - . -. - , - -. . . , . --- -. - . . - -., - --,
LIST OF FIGURES O _Page B-1 EBR-II INSTRUMENTED SUBASSEMBLY XX08. . . . . . . . . . . . . . . B-3 B-2 SUBASSEMBLY XXO8 FUEL BUNDLE INSTRUMENTATION LOADING PLAN . . . . B-6 B-3 RELATIVE XX08 PIN POWERS FOR RUN 97A. . . . . . . . . . . . . . . B-9 B-4 XXO8 SUBASSEMBLY AND FORE-2M AXIAL POWER DISTRIBUTION . . . . . . B-11 B-5 FORE-2M EQUIVALENT THERMAL-HYDRAULIC CHANNEL MODEL. . . . . . . . B 13 B-6 XXO8 SUBASSEMBLY CORNER-CORNER TEMPERATURE DISTRIBUTION . . . . . B-16 B-7 XX08 SUBASSEMBLY FLAT-FLAT TEMPERATURE DISTRIBUTION . . . . . . . B-17 B-8 FLOW MALDISTRIBUTION FACTOR (FMD) . . . . . . . . . . . . . . . . B-18 B-9 COOLANT TEMPERATURE COMPARIS0N FORE-2M, COBRA, XX08 . . . . . . . B.20 B-10 XX08 NEAR CENTER PINS AND FORE-2M C0OLANT TEMPERATURE COM-PARIS 0N FOR AVERAGE CHANNEL LOCATION Z = 5.5 INCH . . . . . . . . B-22 3-11 XX08 NEAR CENTEP PIN AND FORE-2M COOLANT TEMPERATURE COM-PARIS 0N FOR AVERAGE CHANNEL LOCATION Z = 9.5 INCH . . . . . . . . B-22 B-12 XX08 NEAR CENTER PINS AND FORE-2M COOLANT TEMPERATURE COM-PARIS 0N FOR AVERAGE CHANNEL LOCATION Z = 12.68 INCH . . . . . . . B-24 B-13 XX08 (TTC AVERAGE) AND FORE-2M COOLANT TEMPERATURE COMPARISON FOR AVERAGE CHANNEL LOCATION Z = 12.68 INCH . . . . . . . . . . . B-24 B-14 XX08 ELEMENT 16 AND FORE-2M COOLANT TEMPERATURE COMPARIS0N AT LOCATION Z = 12.68 INCHES (CASE 7) . . . . . . . . . . . . . . B-26 B-15 XXO8 ELEMENT 31 AND FORE-2M COOLANT TEMEERATURE COMPARISON AT LOCATION Z = 12.68 INCHES (CASE 7) . . . . . . . . . . . . . . B-26 1 O B-ii
LIST OF FIGURES (Cont'd) O V B-16 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION P.agg Z = 5.5 INCHES FOR CASES 2 THROUGH 4. . . . . . . . . . . . . . . B-27 B-17 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARIS0N AT LOCATION Z = 9.5 INCHES FOR CASES 2 THROUGH 4. . . . . . . . . . . . . . . B-27 B-18 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 12.68 INCHES FOR CASES 2 THROUGH 4. . . . . . . . . . . . . . B-29 B-19 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARIS0N AT LOCATION Z = 20.25 INCHES FOR CASES 2 THROUGH 4. . . . . . . . . . . . . . B-29 B-20 XX08 ELEMENT 10 AND FORE-2M COOLANT TEMPERATURE COMAPRIS0N AT LOCATION Z = 5.5 INCHES (CASE 5) . . . . . . . . . . . . . . . B-30 B-21 XX08 ELEMENT 23 AND FORE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 5.5 INCHES (CASE 5) . . . . . . . . . . . . . . . B-30 B-22 XX08 ELEMENT 17 AND F0RE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 9.5 INCHES (CASE 6) . . . . . . . , , , . . . . . B-32 B-23 XX08 ELEMENT 32 AND F0RE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 9.5 INCHES (CASE 6) . . . . . . . . . . . . . . . B-32 3-24 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARIS0N AT LOCATION Z = 5.5 INCHES, WITH FORE-2M MODEL CONTAINING VARIABLE FLOW MALDISTRIBUTION FACTOR (CASES 3 and 8). . . . . . . . . . . . . . B-33 J-25 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARIS0N AT LOCATION Z = 9.5 INCHES, WITH FORE-2M MODEL CONTAINING VARIABLE FLOW MALDISTRIBUTION FACTOR (CASES 3 and 8). . . . . . . . . . . . . B-34 B-26 XX08 AND FORE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 12.68 INCHES, WITH FORE-2M MODEL CONTAINING VARIABLE FLOW MALDISTRIBUTION FACTOR (CASES 3 and 8). . . . . . . . . . . . . . B-36 8-27 XX08 AND F9RE-2M COOLANT TEMPERATURE COMPARISON AT LOCATION Z = 5.5 INCHES, WITH FORE-2M MODEL CONTAINING VARIABLE FLOW MALDISTRIBUTION FACTOR (CASES 3 and 8) . . . . . . . . . . . . . B-37 B-iii
LIST OF TABLES Page B-1 XX08 SUBASSEMBLY CHARACTERISTICS. . . . . . . . . . . . . . . B-5 B-2 RELATIVE AXIAL POWER DISTRIBUTION FACTORS, F , FOR A a i 24-INCH ELEMENT . . . . . . . . . . . . . . . . . . . . . . . B-8 l B-3 XXO8 SUBASSEMBLY PROPERTIES . . . . . . . . . . . . . . . . . B-10 B-4 DESCRIPTION OF FORE-2M TEST 7A NATURAL CIRCULATION TRANSIENT CASES . . . . . . . . . . . . . . . . . . . . . . . . . . . B-21 0 ( l O B-iv
1.0 INTRODUCTION
~~/ B-1 This report presents comparisons of FORE-2M computer coce[- -]results with EBR-II natural convection test data. Also included is a brief description of the EBR-II natural convection Test 7A, (used for comparison with F9RE-2M code results), the FORE-2M analytical models, assumptions andiinput data used to simulate the XX08 subassembly thermal response under various power and flow conditions for natural convection transients- The' overall FORE-2M issults are in good agreement with the EBR-II Test 7A data [B-2] ,
2.0 TEST 7A DESCRIPTION Test 7A was conducted on September 30, 1978 at the end of EBR-II run 97A. Prior to run 97A, the reactor had been shut down for sfour days. For Test 7A, the reactor power was increased to.17.1 mw (28.5% of rated power), and the flow was 2700 gpm (33% of rated flow). The reactor was held at these conditions for %3 hours. The auxiliary pump, which is normally kept running to insure coolant flow through the core, was then turned off and () the primary flow dropped to 2621 gpm (32.1% of rated flow) normal flow conditions. After 9 minutes, the primary pumps were tripped and about 1.8 - 2.3 seconds after the pump trip, a low flow trip caused a reactor scram. Following the initial flow coastdown, the entire flow in the primary system was the result of natural convection. The test was terminated 19 minutes after the primary pumps were tripped. However, the peak coolant temperatures occurred about 1 minute into the transient, and then gradually decreased until the test was terminated. The overall time period of interest for natural convection analysis is the first two minutes after the primary pump trip. I i i B-1
L 3.0 THE XX08 SUBASSEMBLY DESIGN - The XX08 design is that of an instrumented fueled subassembly located in a O converted control rod position. The entire XX08 subassembly configuration and shape is shown in Figure B-1 A cross-section of the subassembly is shown in Figure B-2 and illustrates the location and types of instrumentation employed. The subassembly instrumentation included two inlet flow meters, six fuel centerline thermocouples, two inlet and three outlet coolant thermo-couples, two coolant thermocouples at 0.4 and 0.7 of the core height (24 inches) positions , and nine coolant thermocouples near the top of core. The nine latter thermocouples were placed approximately 0.85 inches down from the top of the fuel position so that they would be at the same axial level as the fuel centerline thermocouples. All of the elements in the subassembly were fueled except for three: two elements were used as conduits for the flow meter leads which passed through the fueled region of the subassembly, and one contained two rhodium self-powered detectors (SPD). The XX08 subassembly contains 61 elements, each with an outer diameter of 0.174 inch, which were spirally wrapped with a 0.049 inch spacer wire utilizing a 6.0 inch helical pitch. Of the 61 elements, 58 were fueled with a uranium metal - 5 wt % fission alloy designated as the EBR-II Mark-II driver fuel. This design utilizes a U 235 enrichment of 66.7% with a fuel content of approxi-mately 52 g/ pin. The fuel pellet diameter was 0.130 inches and the gap between the pellet and cladding was filled with sodium. The cladding and sodium bond thicknesses were 0.012 inch and 0.010 inch, respectively. The interior dimension of the inner hex can was 1.83 inch across flats and the hex can wall was 0.040 inch thick. The XXO8 subassembly sits in a converted control rod position which, in the core region, consists of a hexagonal guide tube which is also 0.040 inch thick. This placement inside a hexagonal guide tube creates the effect of a double hex duct in which a sodium filled gap, or thimble region, 0.150 inch thick is formed. A summary of these dimensions is given in Table B-1. (*)The active core region is 13.5 inches long. B-2
O SHEATHED LEADS CENTER OF CORE 2708 ELEMENT SUPPORT GRIO CENTER ,1r- E N D 08 : 1808 0F 1 J. FITTING 758 CORE (Z=08) C
"' % COUNTER-
.' k - FLOW CLOCKWISE
. N OlFFUSER ' '
UPPER ADAPTER 908
-COOLANT OUTLETS TOP VIEW ~~
WIRE WRAP
, ORIENTATION FTC
* -- UPPER FLOWMETER i
l -a " CENTER o
^
[ ITh I
- OUTLET OF THERMOCOUPLES CORE -"
i i hO NO e--LOWER 8' 0000% FLOWMETER - FLOW MIXER SECTION A A SHOWING FUEL CENTERLINE 10" THERMOCOUPLES (FTC) I 2
.l .
r END FITTihG d , , n-- FUEL ELEMENTS (/ ,e UPPER ADAPTER _Q(s L
; , 24" lo8l-C00LANT r HEXAGONAL j oo INLET p TUBE l n
93 37/64" 13.5" CORE < ! n
" *-LOWER ADAPTER
. rC00LANT y p l INLET 0.50" P l "
FLOWER ADAPTER g j FUEL U I , ELEMENT I Figure B-l. EBR-II Instrumented Subassembly XX08 8021-29 ! B-3 l
The total length of the fuel elements is 24.0 inches. This length is comprised of a 0.5 inch coolant entry region (containing a stainless steel grid support attachment, a 13.5 inch fueled region, and a 10.0 inch gas plenum space (of which 1.5 inch is filled with sodium). The wire wrap is attached at the bottom of the fuel element (Z = 0 inch) in the 75* position (the O' position is toward the core center) and spirals in a counterclockwise direction. The wire wrap rotates 60 per inch and at the bottom-of-core position is at the 105' mark (Z = 0.5 inch). Additional details are avail-able in Table B-1 and Figure B-2 O i 3-4 u
I i i TABLE B-1 XXO8 SUBASSEMBLY CHARACTERISTICS [8-2] Y Subassembly Parameters Dimension L Hex can, flat-to-flat I.D. 1.83 in. F
', Hex can wall 0.04 in.
6 Thimble gap (between hex valls) 0.15 in.
! . Fuel Pellet Diameter 0.D. 0.13 in.
Fuel Pin Diameter 0.D. 0.174 in. [ Wire wrap diameter 0.049 in. { Wire wrap pitch 6.0 in. . Sodium bond thickness 0.010 in. l i Clad thickness 0.012 in. i ' Number of elements 61 l Wire wrap zero reference angle l
; (start of fuel, Z = 0.5") 105*
J q g Steady State Pretest Conditions Value 3
- Total bundle flow, ib/hr 5.804 x 10 Total thimble flow, ib/hr 0.6052 x 10 I
Average element heat flux (over 5 24 in. of length) Beu/hr-fc2 0.67615 x 10 l l Inlet temperature, 'F 664 Inner Hex can average heat flux, Btu / 3 hr-ft 2** (over 24 in. of length) 1.69 x 10 1 1 Total subassembly power, kW 110.1 i Total S/A pressure drop (inlet and I outlet orifices included), psi 3. 7-I l Auxiliary pump has been turned off. l It was assumed that this heat flux was equally divided over both l l inner (bundle side) and outer (thimble side) hex can faces.
g CORE
.NTER ,
Q@@OQQ ,aM. gggg g FOS,T,eN
@Q@@@@@QQ QQ@QQQQQ SUBASS BLY THIMBLE THIMBLE FLOW LEGEND G
I FM FLOWMETER LEAD CONDUlT 2 SPD SELF POWERED NEUTRON FLUX DETECTOR 2 l BTC COOLANT TC CORE BOTTOM LEVEL 2 4TC COOLANT TC 0.137M ABOVE CORE BOTTOM LEVEL 2 7TC COOLANT TC 0.240M ABOVE CORE BOTTOM LEVEL 2 TTC COOLANT TC 0.322M ABOVE CORE BOTTOM LEVEL 9 i ( 15TC COOLANT TC 0.514M ABOVE CORE BOTTOM LEVEL 1
- OTC COOLANT TC 0.033M ABOVE CORE BOTTOM LEVEL 2 FTC FUEL CENTERLINE TC 0.322M ABOVE CORE BOTTOM LEVEL 6 EL EXTENDED 0.66M LENGTH MK -Il FUEL ELEMENTS 6 NOTES
- 1. THE OUTLET COOLANT TCs(NOT SHOWN) ARE LOCATED NEAR THE SUBASSEMBLY COOLANT EXIT
- 2. THE DOTS REPRESENTS THE LATERAL POSITIONS OF THE TC JUNCTIONS IRRESPECTIVE OF AXIAL LOCATION Figure 11-2. Suliassemlity XX0M Fuel Ilumile Instrumentation Loading Plan 8021 B-6
4.0 EBR-II TEST 7A DATA XX08 SUBASSEMBLY O The steady state pretest flow conditions in the XX08 subassembly were: 665'F inlet temperature, 5804 lb/hr bundle flow and 110.1 kw power. Th
~
relative pin power for each of the XX08 elements is shown in Figure B-3 . The relative axial power profile tabulated in Table B-2 is shown in Figure B-4 along with F9RE-2M approximation. Table 3-3 contains a list of the element material property drta for the fuel, cladding and steel. O i h 4 3-7
., TABLE B-2 .
1 -
~.
~ -
Relative Axial Pever Disteibution Factors, F , for a 24 in. Element l
, F, .for.X263B q Z Z/L Relative Power Factor, Fa Gamma Heating Only 0 0 0.017 0.835 0.5 0.020 0.020 f . 0.992 l 0.501 0.021 1.641 0.993 1 0.042 1.650 1.150 2 0.083 1.706 1.354 3 0.125 1.784 1.499 4 0.167 1.853 1.590 5 0.208 1.909 1.653 6 0.250 1.931 1.681 7 0.292 1.939 1.698 8 0.333 1.920 1.690 9 0.375 1.877 1.673 10' O.417
- 1.810 1.622 11 0.458 1.713 1.555 12 0.500 1.610 . 1.426 13 0.542 1.508 1.273 14 0.583 1.414 0.992 14.01 0.584 0.020 0.992 16 0.667 0.010 0.508 18 0.750 0.006 ' 0.293 20 0.833 0.004 0.169 24.0 1.000 0.001 0.068 i
e Note: The integral of these values over the 24 in. cora length is "one."
~, . . B-8 k
l I
@@@@@@@@@@@@@ ii
!=
"@@@@@@@@@@@@@@@@@/y k
k . ..... i
=
k k_ o k k k o
~
1 O O O
O TABLE B-3 XXO8 PROPERTY TABLE - PROPERTY FUEL THERMAL CONDUCTIVITY, Btu /hr-ft- F 12.9 SPECIFIC HEAT, Btu /lb- F 0.0445 DENSITY, lb/ft 3 972 316 STEEL THERMAL CONDUCTIVITY, Btu /hr-ft *F 12.0 SPECIFIC HEAT, Btu /lb- F 0.14 DENSITY,lb/ft 481.0 CLADDING THICKNESS, in. 0.012 BOND BOND HEAT TRANSFER COEFFICIENT, Btu /hr-ft 2 *F 9999.0 l B-10
2.2 RELATIVE AX1AL POWER DISTRIBUTION FACTORS, Fa!E'2I 2.1 - FWRE2M
---- EBR TEST 7A DATAIB'2l 2.0 -
, - ~ ,
y 1.9 -
'g '
d p / \
$ 1.8 -
f \
$ / '.
E / \ 2 '
/ \
g 1.7 -
, \
p I /
< \
d I \ 1.6 s s I \ I \ s \ 1.5 4 '
\ l 8
\
8
\
I
\
s g 1.4 + i l i
/: '
/
I I I I I I l O.0 l 0 2 4 6 8 10 12 to 16 AXIAL LENGTH (INCHES) i i l Figure B4. XX08 Subassembly and FORE-2M Axial Power Distribution l 8021-19 B-11
5.0 FORE-2M PIN / CHANNEL MODEL The FORE-2M code solves steady state and transient thermal-hydrualic reactor core performance. The analytical model chosen to represent an XXO8 pin channel for comparison of the FORE-2M code with EBR-II (XX08 subassembly) test data consists of an equivalent cylindrical thermal-hydraulic channel as shown in Figure B-5. The key core parameter data used in generating the model is listed in Table B-3. The XX08 fuel pin consists of 13.5 inches of fuel in the active region, and contains a 10.0 inch gas plenum space above and a 0.5 inch space below the fueled region for a total element length of 24.0 inches. ' The FORE-2M model is divided into seven axial nodes and five radial nodes as illustrated in Figure B-5 Six axial nodes are used to model the active core region (fueled portion), and the last (node 7) axial node is used to represent the gas plenum region. The inlet region (0.5 inch of element) below the active region is neglected and the small power generation is lumped into the first axial node. The radial nodalization consists of three (3) equal volume fuel nodes. 1 cladding node with sodium bonded gap and a coolant node. Earli studiesEB-Nshow the same results as with 7' radial node fuel model. The wire wrap and duct masses are coupled via lumping mass and surface area,as described in Reference [B-1 ]. The axial power distribution data was taken from data given in Reference 2, and is in Figure 4.2. which also shows the axial nodal approximation used in the FORE-2M code [B ,1] As shown, the FORE-2M model data has constant power per each axial node and each axial node is sized based on relative constant power distribution in the node as well as close proximity to the axial location of the test thermocouple placement in the XX08 subassembly fuel elements. The calculations for heat generation and heat transfer was based on the assump-tion that the fuel pin was of one material (fuel) along the entire length. Therefore, properties such as conductivity, density, specific heat and also gap coefficients are held constant and equal to those given in Table B-3. O B-12
O n
*= $
5 i- o 13 y .! E l5t E
. 2 zh y d 2 .l.. 52 5 !! 8 a
la le E g g u o l v a n 6 -- I
. n M I . . .
I m u In 4 5 -EQUIVALENT AVERAGE S * * *l
- E CHANNEL BOUNDARY I
a t. $ - SODIUM BOND
- l dS b . . .. h Je '
N COOLANT h, m-l 4 [ Q
~
FLOW CHANNEL in 2 .., .. 4 * * *
- H + - -t '
FUEL T 1 l
*i *l * ,
,,, \N /
CtA0 Figure B-5. FORE-251 Equivalent Thermal-llydraulic Channel Stodel 8021-15 B-13
l However, to account for tf e gas plenum region of the element (last 10 inches the sodium bond heat transfer coefficient was reduced by a factor of 10 4 heat generation rate set to zero. All sodium properties varied as a function of temperature according to property data contained in the FORE-2M . compute.r codeEB-l3 Pin geometry and physical dimensions such as length, thickness, area, gap sizes remained constant throughout the thermal transients, and are listed ( in Table c.1. 6 Total average pin power from Reference 3-2and mass velocity are input as functions of time. The effective flow area for the mass velocity calculation is calculated using the average (XX08 subassembly) flow rate. The flow in some cases contains a flow maldistribution factor (FMD). In cases where flow maldistribution is considered, the flow rate is divided by the flow maldistribution factor (FMD) calculated by using the COBRA codeEB-5] The nominal conditions of 28.1% power (110.1 kw) and 32.1% flow (5804 lb/hr) were used to establish initial conditions for the Test 7A comparison. Observation of the data in ReferenceB-Sihows a rather constant FMD, at higher steady state flows, and noting in Figure D-4: that the subassembly radial power distribution is relatively constant near the subassembly centerline lead to the assumption that the four central elements and their measurements can be averaged. This "near center" pin, therefore, consists of the average power and flow of the four elements near the subassembly center which are labeled in Figure B-2. as 4TC-23, 7T-32,15TC-40 and TTC-31. With a flow maldistribution factor of 1.23 , the FORE-2M pin model becomes equivalent to the "near center" fuel pins. To detennine the effect of variable flow maldistribution, the input transient data profile was altered to reflect a decrease in flow maldistribution (i.e., increased channel flow rate as total assembly flow decreased to natural - convective flow rate levels (<5% total flow). B-14
To simulate the effects of constant and/or variable flow maldistribution, ReferenceB-6shows that when the buoyancy effects are accounted for, the flow maldistribution factor decreases as flow decreases. The overall effect is to distribute more flow to the " hotter" (higher temperature) channels thereby flattening the temperature profile in the subassembly. In addition, radial heat transfer also tends to flatten the temperature profile. This is observed in the EBR-II Test 7A data, which is illustrated in Figures B-6 and Ei-7. The figures show the flat-to-flat coolant temperature profile at the top of the j core (TTC's) and corner-to-corner coolant temperature profile measurements. I Each figure contains the initial coolant temperature (steady state) prior to the pump trip along with the subassembly profile at the time the minimum and maximum temperature peaks occur. As observed, the flat-to-flat and corner-to-corner profiles at minimum and maximum temperatures are extremely flat in comparison to their respective steady state temperature profile. i However, since FORE-2M presently does not have the capability to account for inter- or intra-subassembly radial heat transfer, the comparison of F9RE-2M calculations with EBR-II data considered only the effect of flow distribution which was simulated in the FORE-2M code via changes in (transient) flow input j data. In one case, this was accomplished by reducing the flow maldistribution factor from 1.23 to 1.0 when flow reached about 20% of design flow (58% of , 32.5% flow). The 20% value is arbitrary and was chosen for the purpose of aemonstrating the effect of a variable flow maldistribution factor in the - i natural convectico transient. For FMD case comparison of F9RE-2M and EBR-II data, the COBRA codeNas used to calculate the channel FMD under quasi-steady state conditions. The C98RA code data used EBR-II Test 7A average 1 power and flow as input to generate the flow maldistribution curve shown in Figure B-8. Each data point represents a COBRA code steady state calculation for FMD. l i )
l t l i o u 940 h 00 ~ INITIAL TEMPE RATURE (STE ADY STATE) C
=
860 E
- s
=
h 820 - ca o 2 ct a AT TIME OF PEAK TEMPERATURE
; g 780 -
o p
----~%
740 - AT TIME OF MINIMUM TEMPERATURE 700 i ______ TTC 5 TTC-16 TTC 31 TTC46 TTC-57 (CORNER) (CORE CENTER) (CORNER) Figure 11-6. Element intra-Subassembly Temperature Distribution as a Function of Radial Position Corner to CornerlII'2I g 9 8
_ . ~ - _ _ _ . . . 1 oc 8 MG ( - 1 INITIAL TEMPERATURE (STEADY STATE) 900 - o-
. 860 -
- i w
g i
; :3 i i >=
< I ec i w
w B20 l co 2 , 1 N 3 AT TIME OF PEAK TEMPERATURE o 780 - g - --- p - . - 740 - 2 AT TIME OF MINIMUM TEMPERATURE I
----___.-------,~~-r------...
l l ! TTC 12 TTC-21 TTC 31 TTC41 TTC-50 I (WALL) (CENTFR LINE) (WALU i 4 Figure H-7. Element intra-Subassembly Temperature Distribution as a Function of Radial Position Flat to Flat lH-2l 1 4 1
l i 1.25 O\ 1000 - NOTE: XXOO RUN 7A BUNDLE AVE. COOLANT . ty. 8 *. TEMPERATURE AND CHANNEL FMD AT I
'O, TOP DF ACTIVE CORE
*s -
1.20
\ FLOW MALDISTRIBUTION FACTOR _
T z
\ A e
\ s 900 g -
1.15 9i o[ - g z e
=
\ > S
* \ i N \ R y ) - 1.10 y s
w I '
- z g ,e---- _ -_._._ g $ \ /
3 \ s' e 5 ** - 1.05 g 800 8 3 4 TEMPERATURE $
=
. 5 1'00
- 700 e'* */ l I I I I i
0.40 0 20 40 60 80 100 120 TIME (SEC) Figure 11-8. Flow Maldistribution Factor (FMDilII-S I 8021-18 O B-18
6.0 FORE-2M RESULTS Observation of the wide spread of test data temperature measurements in the XXO8' subassembly caused concern for steady state axial temperature profile in the' FORE-2M code. Therefore, as a check, the average power and flow from the XX08 subassembly data was input to COBRA code ~ to determine the axial temperature profile and the average FMD factor for the "near center" pins. The COBRA results are compared to the FORE-2M calculations with an FMD = 1.23; and the EBR-II (XX08) Test 7A data measurements located near the sub-assembly center and are shown in Figure B-9 As shown, the FORE-2M and COBRA codes are in good agreement (within 5'F) of each other, and are within 25 F of the test data. Eight transient cases were made using the FORE-2M code and compared to EBR-II (XX08 subassembly) Test 7A data. Each case represents either a variation in the pin model (i.e., a near center pin) or a p~arametric change (i.e., FMD). Table B-4 contains a brief description of each case. The purpose of the transients was to determine the sensitivity of flow maldistribution, radial location of subassembly pin and reactor. scram time. CASE 1 The FORE-2M nodel consists of the power and flow input data calculated on the basis of using the subassembly average flow and power (FMD = 1.0). Since no one pin contains all axial temperature measurements, TTC measurements 4TC-23, 7TC-32, 15TC-40 and TTC-31 are assumed to represent an average channel. In this case, it is assumed the average pin corresponds to the pin position near the subassembly center. Also, for worst case comparison, the scram time = 2.3 seconds. Figures B-10throughB-13 show the comparison of the FORE-2M average pin calculations with EBR-II Test 7A data. Figure C-10 shows the FORE-2M average channel coolant temperature compared to EBR-II data measurement 4TC-23. As shown, the FORE-2M steady state temperature is about 25 F below the test data. However, under transient conditions, the FORE-2M result is within a few degrees (<5*F or 9% of maximum temperature rise), and the minimum and maximum temperature peaks occur at the same time. Figure 3-11 shows the FORE-2M average channel coolant temperature at the 9.5 inch location compared to EBR-II measurement 7TC-32. Again, the initial temperature is about 25 F below the test data and the transient data is within 5'F (or 6% of maximum temperature rise). B-19
940 900 - A TTC 860 - C
~
w 7TC 5 g 820 - 5 s N A
$ 780 - 4TC 3
8 u 740 - CO BRA IV CALCULATIONS o (CHANNEL #3) a XXO8 TEST DATA (NEAR ASSEMBLY CENTER) BTC,4TC,7TC,TTC AS SHOWN IN FIGURE B-2 700 - O FWRE 2M CALCULATIONS a O (1.23 FMD NOMINAL) 660 0 2 4 6 8 10 12 14 AXlAL -INCHES ABOVE CORE BOTTOM Figure 11-9. FORE-2M and Cobra Code Comparson of Axial Temperature Distributions With EllR-il Test Data (7A) h021 -4 O B-20
i c ; t , TABLE B-4 FORE-2M TEST 7A NATURAL CONVECTION fO TRANSIENT CASES DESCRIPTION CASE NUMBER DESCRIPTION 1 Average channel. Pin power and flow average of sub-assembly total flow maldistribution factor (FMD) = 1.0. 2 Near center of subassembly. Power and flow based on average of radial power factor and average channel flow, maldistribution factor equal a constant. . FMD = 1.23; scram at 2.3 seconds. 3 Same as Case 2 except that flow maldistribution factor step changed to one (1) at flow rates less than 20% of design flow. 1.0 < FMD < 1.23; scram at 2.3 seconds. 4 Same as Case 3 except reactor scram occurs at 1.8 seconds after pump trip. 5 Flow, FMD factor, and power calculated as per pin position 10. TTC = 5.4 in. above core bottom. O- 6 Flow, FMD factor and power calculated as per pin position 17. TTC = 9.5 in. above core bottom. 7 Flow, FMD factor and power calculated as per pin position 16. TTC = 12.68 in, above core bottom. 8 Same as Case 2 except FMD = F(T) per Figure B-8. c.91 m)
e
-g 2 FIGUREB-10 xxca No TEWP 0.137m A80VE CCRC 80TTOW. 4TC23 .5
. l EBR R Test 7A. PRlWARY SYSTEu
=
. 'I CASE 1 0F TABLE 34 o FORE-2M AVG. CHANNEL MODEL 5
EBR-II TEST DATA
.e.
3 , 3 , t: *u h *
*g eI - .
,c.. ,
, 3=
o oo. .*. M p A
= -
p w_ *
.g g -
A o u,
~
M "* I E -r s -
- to O C 20 30 , 40 50 to 70 80 90 90 te 12 0
. TIME, seconds 1
- e. .
0 g FIGURE B-lla XxCS Na TEMP 0.240m ABOVE CORE BOTTOW. 77C32 .g
= l EBR R Test 7A. PR! WARY SYSTEl4 3
3 "k CASE 10F TABLE B-4 o FORE-2M AVG. CHANNEL MODEL .g l , . EBR-II TEST DATA .: 2 s wg o "T u
.. oo k *
*g et .- _m 0
l i i a
~
.Am l
6
/ -
^
l*
-i l
- - v - .g
.o I .*
-s l
-4 0 0 20 30 40 50 60 70 to 90 90 t9 t23 1 7%C. soccecs B-22
l l I l Figure B-12 shows the FORE-2M average coolant temperature at 12.5 inches compared to EBR-II data measurement TTC-21. Again, the FORE-2M initial temperature is about 25'F below test data. Under transient conditions, the FORE-2M minimum temperature is 5 - 10"F (or 13% of the temperature rise) below the test data, and the maximum temperature is about 20*F (or 20% of the maximum temperature rise) higher than the test data. Figure B-13 shows the FORE-2M average coolant temperature at 12.5 inches compared to the average of all TTC's in EBR-II test measurements. As shown, the initial temperature agrees very well (within 5*F) and the minimum tem-peratures are within 5*F. However, the maximum temperature is about 20-25 F (or 26% of maximum temperature rise) higher than the test data. CASE 2 To account for flow maldistribution and relative radial power variation in the XX08 subassembly, the "near center" region (consisting of temperature measure-ments 4TC-23, 7TC-32,15TC-40, TTC-31 and FTC-31) was chosen for comparison with the FORE-2M code. The relative power was calculated by averaging the relative power of the "near center" four pins (Figure 8-3 ) and the flow maldistribution factor (FMD) was calculated using COBRA code data . The flow maldistribution factor was held constant (FMD = 1.23) throughout the transient. CASE 3 The FORE-2M model data input was changed to simulate a change in the flow maldistribution factor. This was accomplished by changing the average sub-assembly flow profile to account for a decrease in FMD. The FMD was (step) changed from 1.23 to 1.0 after the subassembly flow. dropped below 20% of design flow rate. CASE 4 The scram time is changed to 1.8 seconds from 2.3 seconds to determine the sensitivity of thermal response to scram time. fm ( B-23
$ ~$ .
FIGURE B-12 xxos No TEVP 0.120m A80VE CCRE 80TTCu, TTC-21 .tr I j EBR N Test 7A. PRIWARY SYSTEW
- g CASE 1 0F TABLE B-l $
g v s,. o FORE-2M AVG. CHANNEL MODEL .g a O
- EBR-II TEST DATA .
w 2, ov
'E 1
-2
- h 9
N -t
- O .R
- N
, O O e-9 0
~
m ~ Lk \ y 3 g . x I
-r l E -E i a :
j - 80 5 9 2b b 40 Sh 40 70 40 m 90 f10 12 0 t T1WE. seconds
* -2
. a g FIGURE B-l'3 xxca mERAcE TEWPERATURE OF ALL TTC's .!r
. l EBR 11 Test 7A. PRlWARY SYSTEW y .. .g CASE 10F TABLE 34 o FORE-2M AVG. CHANNEL MODEL 1 5-EBR-II TEST DATA .g e
* - oi 5
1 ._ sg ~" ' u I
.x -
'I el O
.a* =
O O O _
.g g ,
"2 1
-r e
s* 5 s
.e 4 e a m 4 a m a m a = m T1WE. seconds -
B-24
Figures 3-14 through 3-18 show the comparison of FORE-2M Cases 2 through 4 temperature calculations with EBR-II (XXO8) Test 7A data. The FORE-2M steady state temperatures are about 25 F higher than test data. During transient conditions, the FORE-2M calculations are within 20"F except when the maximum peak temperatures occur. The fixed FMD (Case 2) has a temperature difference of about 40*F, whfe.h is higher than the variable FMD (Cases 3 and 4) which are only 20*F higher than the test data. Figure B-141s the comparison of FORE-2M and Test 7A coolant temperatures at 5.5 inches above the core bottom. As shown, the initial temperatures are within 5*F for all three cases. Also, Case 2 (the fixed FMD case) has the largest temperature difference (about 15*F or 14% of maximum temperature rise) at the maximum temperature condition. Cases 3 and 4 (FMD decreasing from 1.23 to 1.0) show an insensitivity to a small (0.5 second) variation in scram time. In all cases, the transient temperature behavior is in excellent agree-ment (less than 5 F or 5% of maximum temperature rise) with the EBR-II test data. Figure B-15 shows the comparison of FORE-2M Cases 2, 3 and 4 and Test 7A data coolant temperatures for the measured location 9.5 inches above the core k bottom. Similarly, as in the above comparison, the constant FMD (Case 2) does not agree as well as the variable FMD (Cases 3 and 4). Again, the steady states are within 5 F of each other. The variable FMD case is within 5*F (or 6% of maximum temperature rise) of the test data, while the constant FMD case (FMD = 1.23) is 25*F (or 29% of maximum temperature rise) at the time of maximum temperature. Figure B-16 shows the comparison of the three FORE-2M cases and EBR-II Test 7A data coolant temperatures corresponding to measurements at a location 12.65 inch above the core bottom. The FORE-2M steady state temperature in all cases is about 15-20*F (or 18-24% of maximum temperature rise) higher than the test data. Likewise, the maximum difference is 15-20 F higher at maximum peak temperature for the variable FMD (Cases 3 and 4), while the fixed FMD (Case 2) is about 35 F (or 35% of maximum temperature rise) higher. However, when considering the steady state offset, these differences reduce to less than 5*F and 15*F, respectively, during transient conditions. B-25
E -
-5 .
FIGURE D-14I xxca No TEMP Q.137m ABOVE CCRC 80TTCW. 4TC23 .2 I l EBR ll Test 7A. PRlWARY SYSTEM
- g -
I o FORE-2M CASE 2 FMD = 1.23 (SCRAM = 2.3) g I c* 0 FORE-2M CASE 3 FMD STEPPED 1.23 TO 1.0 (SCRAM = 2.3) ,
.e l %
A 5 FORE-2M CASE 4 FMD STEPPED 1.23 TO 1.0 (SCRAM, = 1.8) ;) EBR-II TEST DATA .
':: *u .
Si' s - Ei
*I :,W
.a u C
~
a
.f
~
oo . a O Mt% O n
- ,W a l
I y- . g- ., e S *
-9 5 9 2h 30 , 40 SO 80 7b 80 90 90 11 0 12 0 DME. seconds l
l [, -E FIGURE B-15 xxca No TEMP O.24cm ABOVE CORE BOTTou. 7TC32 I l EBR n Teer 7A. PRlWARY SYSTEM
.g g "I O q ,g i l
- I , . o l l .
g ,
' SAME AS FIGURE 3-14 g .w=- 5 Sg -
"T u h *
'h m }*
et .z - 2 00 o
- v g g O O '
g
'% e -
< 56
.g _
~-
5 y 5 2 l 1 :
*C 0 4 20 30 4a so e4 70 80 90 10 0 no
- 12 0
. DME. seconds B-26
t
~ . _ . . _ ,-
-2*
I- FIGUEE B-16 xxos No rEup o.32cm AsovE cesE soTToW. rTC-31 .a*
~} EBR al Test 7A. PR: MARY SYSTEM * * * " * - '
2
@ O ',
- . . . . . _ , 1 g 4 _
,3 w w' O 3 .
' SAME AS FIGURE B-17 -
a - i
.g "E - .
l o 0 o
. 3
* *I g 4 g o gd g '
n o I . [ %& - JI E
..Y I
*k ,
3 E
-I
-. s . se se 4 as e's to as as 3
.o as se
_ _ _ _ _ __ M seconds g
. FIGURE B-17:
s xxos Na TEMP o.S!4m A80VE CCRC UCTTCW ISTC40 .s* l EBR 11 Test 7A. PRlWARY SYSTEM e .
# 0
-{
I - FORE-2MCASE2FMD=1.23lSCRAM=2.3) I* W' e
~ O FORE-2M CASE 3 FMD STEPPED 1.23 TO 1.0 (SCRAM i = 2.3)
I a FORE-2M CASE 4 FMD STEPPED 1.23 TO 1.0 (SCRAM I = 1.8) EBR-II TEST DATA
,I 5, I. *
- oO o ',* I .
WI , g o 40 .a' u 4 o O
.A o/
g 3
-t I. - . . 2. e e . s . . I TIME, seconds B-27
Figure B-17 shows the comparison of the three FORE-2M (Cases 2, 3 and 4) com pared to Test 7A data. Note, the FORE-2M calculations correspond to a locatiU F 18.75 inches above the core bottom, while the test data is about 20.25 inches above the core bottom. The FORE-2M result would agree more closely if the calculation represented the 20.25 inch axial location. As can be observed from the test data, the upper portion (non-fueled region) of the pin has a lower temperature than the core exit (active region) because of more radial heat transfer and upper region coolant mixing. FORE-2M computer Cases 5 through 7 are calculations based on precise pin assembly location and thermocouple axial location within/on fuel pins or within the coolant channel region of the fuel pin. Pins labeled 10,16 and 17 in Figure B-1 were selected for comparison with FORE-2M. The flow maldistribution factor was calculated from COBRA data for each pin. The relative radial power factor was adjusted based on values shown in Figure B-3 CASE 5 Case 5 is the FORE-2M model for pin 4TC 10 and axial measurement location at 5.4 inches above the core bottom. Figures B-18 and B-19 show the comparison of FORE-2M (Case 5) temperature results with EBR-II Test 7A data. As shown in Figure B-18, the steady state temperature is 25 F lower in FORE-2M. However, during the transient, FORE-2M predicts a higher (about 28 F or 35% of maximum temperature rise) temperature than the test data. Figure B-19 shows Case 5 prediction compared with element 4TC 23 (near assembly center). In this case, the steady state agrees very well (within 5'F) and the largest tempera-l ture difference is about 15 F (or 19% of the maximum temperature rise) at the time of maximum temperature (at t % 50 seconds). CASE 6 Case 6 is the FORE-2M model of pin 7TC 17 and axial location 9.5 inches above the corc bottom. Figures B-20 and B-21 show the comparison of FORE-2M (Case 6) 1 temperature transient with EBR-II Test 7A data. As shown in Figure B-20, the steady state temperature is about 25 F lower in FORE-2M, but during the transient, the FORE-2M temperature is 25*F (or 25% of maximum temperature rise) higher at maximum temperature. B-?8
{ ~ g' ((gtg( g41 XXQ8 Na TEkP Q.137m AGOVE CCRC DOTTOW. 4TC10 .5 EBR !! Test 7A. PR! WARY SY S T[W * * * * *
. . . . . . .. .. . . .. .. . . . .- . . . . . . . . . . 2 g . . . . . . . . . . . _ . . _._ . 3 a EBR-II TEST DATA
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Figure B-21 shows F9RE-2M (Case 6, which is a near edge pin) compared with test data of a "near center" assembly pin 7TC-32. In this comparison, the steady state temperatures are within 5 F, and the largest temperature 0 difference of about 15 F (or 15% of maximum temperature rise) occurs at time of maximum terperature. CASE 7 Case 7 is the F9RE-2M model of pin FTC 16 and axial location 12.68 inches above the core bottom. Figures B-22 and B-23 show the comparison on FORE-2M (Case 7) center fuel temperature transients, with EBR-II Test 7A data. Figure B-22 shows that the F9RE-2M steady state temperature is about 25 0F lower than the test data, and at the time of the maximum' temperature, the F9RE-2M calculation is about 25 F higher (or 23% of the maximum temperature rise). Figure B-23shows the comparison with "near center" assembly pin FTC 31 with the FORE-2M (Case 7). As shown, the steady state temperature 0 difference is about 25 F lower, an'd the largest temperature difference is 0 about 20 F occurring at the time of maximum temperature. These two compari-sons depict steady state temperature profile skew possibly caused by subchannel flow distribution or bundle distortion as discussed in Reference 8 and also illustrated in Figure B-7. The steady state temperature of PIN 16 and PIN 31 are respectively above and below the ' top hat' temperature profile calculated by CLUSTER and THI3D codes. CASE 8 Case 8 is the "near center" pin FSRE-2M model with variable flow maldistribu-tion factor. The FMD C9 BRA data is input as function of time as shown in Figure B-8, Figures B-24 through B-26 show the comparison of the F9RE-2M results with EBR-II data. Also, included is Case 3 (the stepped change in FMD 1.23 to 1.0). As shown in Figu e B-24, the variable FM) case is about 6 F above the test data and about 5 F (or 7% of maximum temperature rise) above the stepped FMD case. Figure B-25 is a comparison of coolant temperatures at 9.5 inches above core bottom. The variable FMD case is about 5*F above test data and about 8*F_. above the stepped FMD case. O 3-31
! -3 3 FIGURE B-22 xxos ruEL CENTERLINE TEup. rTC-16 3 a EBR II Test 7A. PRIMARY SYSTEu .
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a L, a = m a a 00 s00 - J i I I I I I O 20 40 SO 80 100 120 TIME (SEC) - Figure B-24. F$RE-2M With FMD and EBR-II (XXO8) Test 7A Coolant Temperature Comparison at Axial Location Z = 5.5 Inches I 3
960 a FMD STEPPED FROM 1.23 TO 1.0 AT 10 SEC 920 - O FJRE 2M FMD VARIABLE TEST 7A DATA NOTE: CASF 8 (SEE TABLE B4) 880 - a 840 -. C
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w 5 y 800 - cz: E 3 O 760 - O O N 720 - O a i l l 680 - 640 0 20 40 60 80 100 120 TIME (SEC) Figure B-25. FORE-2M With FMD and EUR-il (XX08) Test 7A Coolant Temperature Comparison at Axial Location Z = 9.5 inches 8021-23 O B-34
i i 4 l Figure B-26 is a comparison of coolant temperatures at 12.68 inches (12.5 inches in F9RE-2H) above the core bottom. Again, the variable FMD case is ! above the test data (sl5*F or 14% of the maximum . temperature rise) and about 5 F above the stepped FMD case. Figure B-27 compares coolant temperatures at 20.5 inches above the core bottom. All calculated coolant-temperatures at this location are higher than ones measured primarily dut ! to radial heat transfer outside of the effective fueled region. ; 4 1 9 i .j i
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720 - o , ao 680 - l 1 l 640 0 20 40 60 80 100 120 TIME (SEC) Figure 11-26. ESRE-2M With FMD and EllR-II (XX08) Test 7A Coolant Temperature Comparison at Axial Location Z = 12.68 Inches 8021-24 O B -36
960 J A 920 - A FMD STEPPED FROM 1.23 TO 1.8 AT 18 SEC 4L 0 F(RE 2M FMD VARIABLE TEST 7A DATA 880 - NOTE: CASE B (SEE T ABLE B4) 840 - C E
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k 760 va O O O O A a 720 - Oo 680 - I ! l l 640 0 20 40 60 80 100 120 TIME (SEC) Figure B-27. FORE-2M With FMD and EBR-II (XX08) Test 7A Coolant Temperature Comparison at Axial Location Z = 20.5 Inches 8021-22 B-37
7.0 C_014CLUSIONS In general, the FORE-2M code compares very well with the EBR-II (XX08 sub-assembly) Test 7A data. The steady state comparison of FORE-2M with the test data agree very well in the " center" of the subassembly. However, at the corners and flats, the difference was about'250 F higher in the core outlet temperature comparison. This is probably due to several factors such as; inter-assembly radial heat transfer which tends to lower the outer (located) pin temperatures. This is observed by noting that the steady state comparison with test data becomes larger as the axial locations approach the core outlet region (i.e., 20 F at 5.5 inch compared to 30 F at 13.5 inch). Also, as mentioned in the Case 7 discussion, abnomal subchannel flow distribution and/or bundle dis-tortion has caused distinct zones of ' cold' and ' warm' subchannel flow. During the natural convection transient conditions, all of the various FORE-2M cases compared very well with the test data. The FORE-2M model which is the average of rods 23, 32, 40 and 31 agreed very well with the "near center" measurement data, and agreed exceptionally well with the average core out-let temperature (9 themocouples averaged). The F9RE-2M result, without effects of variation of flow maldistribution and radial heat transfer, was conservatively high by 20 0F ($20% of maximum temperature rise). With the 0 variable flow maldistribution factor the difference reduces to 10 F (510% ofmaximumtemperaturerise). The largest differences in coolant temperature comparison cases occur when near corner pin measurements were compared with FORE-2M results. The FORE-2M msults were 25-40 F higher than test data as compared to 15-25 F with near center element comparison. This is due to edge channel overcoolir.g. The FORE-2M results are in good agreement and predict conservatively higher temperatures than the EBR-II (XXO8 subassembly) Test 7A data. The FORE-2M results would agree better with the test data if radial heat transfer and flow redistribution were included in the FORE-2M model. A version of FORE-2M, which uses such detailed data as input, is under development as part of the CRBR NCVP. This is demonstrated in the transient comparison using a variable O B-38
O FMD (Case 8) based on C9 BRA code calculated FMD data, which neglected inter-assembly heat transfer but modeled intra-assembly flow and heat redistribution. In this transient case, results compared at the peak temperature increased U from about S F to 15 F from core inlet to outlet, and is probably a more realistic comparison (than other cases). The inter-assembly radial heat transfer affects would be more prominent at the core outlet than at the in-let due to a larger AT plus longer mass transport time to allow heat transfer which would explain the increasing difference between the code prediction and da ta . Finnaly, as previously reported in XXO7 and the FORE-2M comparisons , the power and flow measurement uncertainties are about i 10% and respresent 0 , 10-20 F in temperature deviation. Therefore, the F9RE-2M results are within the region of test measurement uncertainty. O W 9 O B-39
REFERENCES B-1. J. V. Miller and R. D. Coffield, "FBRE-2M: A Modified Version of the FORE-II Computer Program for the Analysis of LMFBR Transients", CRBRP-ARD-0142, November 1976. B-2. J. L. Gillette, et al ., " Compilation of Data from EBR-II Natural Circu-lation Test Procedure EX-140, Section IV, Addendum 7A", ANL/EBR-104, April 1979. B-3. Personal Communication with Paul Betten (ANL) on 1/24/79. B-4. Appendix A of this report. B-5. BNWL-1962, " COBRA-IV-I Interim Version of Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores", March 1976. B-6. R. M. Singer and J. L. Gillette, " Measurements of Subassembly and Core Temperature Distributions in an LMFBR", AIChE Symp. Series, 73,164,
- p. 97,1977.
O B-40 O}}