ML19345B796
| ML19345B796 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/30/1980 |
| From: | Markley R, Tang Y, Vijuk R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19345B780 | List: |
| References | |
| WARD-D-0274, WARD-D-274, NUDOCS 8012020517 | |
| Download: ML19345B796 (400) | |
Text
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ATTACHMENT IV CRBRP NCVP PRE-TEST PREDICTION OF F0TA TEMPERATURES FOR FFTF NATURAL CIRCULATION TEST INITIATED AT 35% POWER AND 75% FLOW CONDITIONS O
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CRBRP NCVP PRE-TEST PREDICTION OF F0TA TEMPERATURES FOR FFTF NATURAL CIRCULATION TEST INITIATED AT 35% POWER AND 75% FLOW by R. D. Coffield K. D. Daschke Y. S. Tang 4
November 1980 REVIEWED:
Y.S[ Tang,AdvisoryEngineer Core T&H Analysis APPROVED:
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R.A.Markley, Manage [
Core T&H A alysis h
R.M.Vijdk, Manager Reactor Analysis and Core Design
1.
Introduction The purpose of this report is to present pre-test predictions of the themal and hydraulic responses of the fueled open test assemblies (F0TA's) in the FTR for the FFTF natural curculation plant startup test initiated at. 35% power and 75% fitw.
The 35% power test is one of the series of natural circulation tests to be performed in the FFTF during acceptance testing.
Pre-test predic-tions for the FFTF 75 and 100% power tests will also be perfomed and documented in a forthcoming report. A summary description of the FTR c.e configuration is given in the appendix.
The n cural circulation tests demonstrate the capability of adequate decay heat removal by means of natural circulation and provide the basis for the verifi-cation of the analytical tools used to predict the flow and temperature responses in LMFBRs.
The COBRA-WC and FORE-2M codes as well as DEMO have been used for the core analyses.
These predictions are the final step in an extensive veri-fication program (Refs.1, 2 and 12) on these codes which includes comparisons with test data, existina analytical solutions, and recognized codes as well as comparisons with numerical solutions and sensitivity analyses. The detailed description and results of these verifications of the COBRA-WC and FORE-2M codes will be published in future reports.
During the transition to and operation in the natural convection cooling mode, the effect of a power-to-flow ratio greater than that at steady state is exper-ienced.
Consequently, cc,re temperatures increase and natural convection phenomena such as inter-and intra-assembly flow redistribution become significant once low flow conditions are reached.
In the CRBR or FTR the core thermal head becomes increa.,ingly significant relative to the form and friction losses across the core for flows below 5% of full flow.
Coupled with the flow redistribution,c significant heat redistribution on an inter-and intra-assembly basis occurs throughout the core due to radial temperature differentials and an increased flow transport time..Both of these effects (i.e., natural convection flow and heat redistribution) have been found to significantly reduce maximum core tem-peratures as demonstrated by the EBR-II natural circulation experiments (Ref. 3) and Brookhaven investigations (Ref. 4).
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2.
ANALYTICAL APPROACH
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To analyze LMFBR decay heat' removal capability, a system of three computer codes has been developed and is used in sequence, i.e.,1) DEM0, for plant-wide analyses (Ref. 5); 2) COBRA-WC, for core system analyses (Ref. 6); and 3) FORE-2M
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for localized core hot rod analyses (Ref. 7). The last two codes provide the two-stage calculational approach for the core analyses under natural circula-tion conditions.
Utilizing DEM0 boundary conditions, a detailed whole-core flow and heat redistribution analyses of all the parallel core assemblies and bypass regions is first performed by the COBRA-WC code. This is done for a sector of symmetry of the reactor core.
By necessity, the calculational matrix or nodal representation, especially distant from the area of interest, is less detailed than that in the area of interest, i.e., F0TA.
Clusters of rods rather than individual subchannels are modeled. The results of this analysis provide the effect of reactor inter-and intra-assembly flow and heat redistribution on the tcmperature response. The data are then used for a detailed analysis on a single, " hot rod" using the FORE-2M code. These latter analyses include effects of localized phenomena and uncertainties in nuclear / thermal-hydraulic / mechanical data.
IV-1
A linkage between the COSRA-WC and FORE-2M codes has been developed and veri-fied to incorporate the inter-and intra-assembly phenomena into the localized hot rod natural circulation analyses (Ref. 8).
For each axial node of the hot rod modeled in F0FE-2M, a heat balance is performed using the expression for the heat transferred to the coolant at that section, Q (X,1) as c
Qc (X)
- Or (X)
- Oex (X)
U) where Qr (x,t) = heat transferred from the rod surface at axial location x and time t; Q,x(x,t) = coolant heat input or loss due to radial conduction and mixing heat transfer and flow redistribution to adjacent coolant channels; directly input from COBRA-WC.
Coupled with this, the axial mass flow rate for each axial node G(x,1) is also input from C C RA-WC analyses.
Both Qex(x,1) and G(X,1) are based on nominal conditions in the COBRA-WC code.
It is conservative to use Qex(x,1) and G(x,t) data in this fashion because these values are lower than those calcu-lt.ted for the hot channel temperature conditions, and thus, result in a cor.servatively higher predicted hot chsnnel temperature.
Boundary conditions for the COBRA-WC and/or FORE-2M analyses (e.g., plenum-to-plenum pressure drop and coolant inlet temperature) are furnished by the plant-wide code, DEM3 (Ref. 9) for several cases: the "best estimate" or nominal case as well as cases with uncertainties.
Likewise, a corresponding modeling of the core parallel flow network, with regard to pressure drop and decay heat uncertainties, can be used in the COBRA-WC analyses for input to the FORE-2M hot rod temperature predictions.
For the 35% test, only the case with nominal conditions and minimum assumed irradiation history was explicitly calculated for both the F0TA 2 and 6 via the DEM0/ COBRA-WC/: 2RE-2M sequential method as an initial base run.
For other cases, the method was used implicitly as will be described later in this section.
Figure 1 shows the model used by the COBRA-WC code for the FFTF F0TA's to c
maintain both a reasonable computer running time and adequate level of detail (Ref.10).
In this case, a 217-rod FFTF fuel assembly has been divided into 37 channels.
Each channel shown is represented by one rod with power and flow conditions corresponding to the average for all rods contained in it.
For the interior, hexagonal shaped regions, this averaged rod would represent a total of 12 rods (other assemblies were modeled in less detail).
To evaluate the effects of core uncertainties, hot channel factors are used (Table 1) in FORE-2M.
Similar to those used in steady state calcula-tions, the direct and statistical type factors are conservatively applied by the semi-statistical method.
For instance, the inlet flow uncertainty of -5%
(Ref. 11) to the FOTA is conservatively assumed to occur.
The statistical subfactors which are statistically combined include those due to uncertainties in the decay heat calculations (+25%), the pressure drop calculational uncer-tainties (specified in Ref. 9) and the factors for power level measurement, i
nuclear power distribution and coolant property uncertainties as used in Ref. 8.
The effect of the recent FFTF shield / orifice assembly pressure loss coefficient data (from HEDL) was accounted for in these predictions.
The procedure for evaluatino such a global effect as well as those due to decay heat and pressure drop calculational uncertainties was to utilize DEMD code results (Appendix A) to determine flow variations. The FORE-2M code was then used to determine resultant temperature changes. Similarly, the effect of the core IV-2 t
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Location A for Row 2 F0TA on Rod (9,9) FFTF Designation Location B for Row 2 FOTA on Rod (8,8) FFTF Designation Location C for Row 6 F0TA on Rod (4,4) FFTF Designation Location D for Row 6 FOTA on Rod (6,6) FFTF Designation FIGURE 1 TYPICAL COBRA-WC 37 CHANilEL MODEL OF A 217-ROD A5SEMBLY i
IV-3 L
TABLE 1 HOT CHANNEL FACTORS APPLIED TO HOT R0D TEMPERATURES IN FORE-2M CALCULATIONS FCR N/C TESTS,
("1 dr-UNC" AND "56 Hr-UNC" CASES)
A.
DIRECT SUBFACTORS 1)
INLET FLOW UNCERTAINTY 1.05 B.
STATISTICAL SUBFACTORS (3a).
- 2) DECAY HEAT UNCERTAINTY EFFECT 1.20*
3)
PRESSURE DROP CALCULATIONAL UNCERTAINTY EFFECT 1.20*
4)
POWER LEVEL MFASUREMENTS 1.08
- 5) COOLANT PROPERTIES 1. 01
- 6) NUCLEARPOWEkDISTRIBUTION 1.035 t
- These factors represent values used at the time when peak temperatures occur during the natural circulation transient.
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- m IV-4
inter-and intra-assembly flow and heat redistribution (Ref. 8) was accounted for via the COBRA-WC/ FORE-2M linkage.
This was done by comparing the sodium temperatures utilizing DEM0/ COBRA-WC/ FORE-2M sequential computations with the same analysis which neglects the COBRA-WC computations for Qex(x,r) and G(x,r).
These results were then applied to the FORE-2M temperature predictions as bias factors for the "56 Hr-NOM" and "UNC" cases (defined in Section 3).
This is a conventional method of accounting for perturbations on a multi-parameter problem.
For natural circulation test predictions, the aforementio.ned effects including the hot channel factors are not constant with time, due to transient effects. The actual time-dependent factors were used to obtain the 3a uncertainty maximum temperatures. Table 1 presents values of these factors used at the time when the peak temperature occurs.
3.
CASE DEFINITION AND B0UNDARY CONDITIONS The boundary conditions of the reactor, such as initial power, inlet and out-let plenum temperatures (as a function of time),. reactor inlet flow and the nozzle-to-nozzle pressure drop (as a function of time) are provided by the system code, DEMO.
Detailed description of COBRA-WC modeling and output pro-vided to the FORE-2M code will be described in a separate report.
Since the 35% power /75% flow initial condition test is the ff rst of the signi-ficant dynamic tests which are part of the FFTF Acceptance Test Procedure, the allowable decay heat after shutdown will most likely be minimized. This is accomplished by only allowing low irradiation histories (burnup) on the core prior to the initiation of the test.
The minimum irradiation time is estimated to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 355 power, and the maximum irradiation time is estimated to be 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> at 35% power.
Thus, these irradiation histories should bracket the actual irradiation time before shutdown. The effect of such variation in decay heat on the temperature response is significant in this case, and will be shown separately. As mentioned in Reference 9, this and some other boundary conditions used in the calculation are assumed conditions.
It is likely that post-test analysis made with decay power based on actual pre-test irradiation history will be necessary.
Ocher boundary conditions such as the pump coast-down characteristic and intermediate loop pony motor operation should also be accounted for in the post-test analyses.
The following cases are therefore analyzed:
Case "I Hr-NOM" - Nominal case with no uncertainties, but includes flow and s
heat redistribution with only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> irradiation history at 35% core power.
Initial base runs were performed with DEMD, COBRA-WC, and FORE-2M; correc-tions were made for the recent shield / orifice assembly pressure loss co-efficient data.
System parameters from Reference 9 were utilized.
Case "56 Hr-NOM" - Nominal case as in "I Hr-NOM" except with 56-hour c
irradiation history at 35% core power.
Runs were performed with DEMD and FORE-2M; correction factors were applied for 1) inter and intra assembly flow and heat distribution and 2) shield / orifice assembly pressure loss coefficient.
System parameters listed in Appendix A were utilized.
Case "56 Hr-UNC" - Case with 3a uncertainties including decay heat, AP n
characteristics, flow and heat redistribution for 56-hour irradiation i
history at 35% core power. Runs were made with DEMD and FDRE-2M; correc-tion factors were applied for 1) inter and intra assembly flow and heat l
distribution and 2) shield / orifice assembly pressure loss coefficient.
System parameters listed in Appendix A and uncertainties lisced in Table 1 l
uere utilized.
IV-5
Case "1 Hr-UNC" - Same case as in "56 Hr-UNC" except with 1-hour irradia-e tion history at 35'; core power.
Correction factors used in the "56 Hr-UNC" case were adjusted to account for 1-hour decay heat isvels and applied to the "1 Hr-NOM" case, Case "CRBRP/ PRE-NCVP" - Current CRBRP conservative design approach (which e
neglects the effects of inter-and intra-assembly heat and flow redistri-bution and combines the hot channel factors conservatively) for 1-hour and 56-hour irradiation history at 35% core power.
DEMD and FORE-2M runs were made utilizing Reference 9 system parameters for the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> case and Apper dix A system parameters for the 56 hour6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> case.
Shutdowr power data for the 1-hour irradiat' ion cases are provided in Reference 9.
The correspondence data for the 56-hour cases are given in Appendix A.
As shown in Figure 1, the F0TA is simulated by the COBRA-WC code by a 37-
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channel model.
For the Row 2 F0TA, channels 9,10 and 15 are of the most interest, with channel 15 representing the hottest channel in this assembly with a very nearly flat power profile. Because of the nuclear flux distribu-tion and heat transfer to the surrounding assembly of the Row 6 F0TA, channels 8,13 and 15 are of most interest, with channel 13 representing the hottest channel.
Other initial conditions are (Ref. 9 ):
e Inlet temperature:
631*F e Nozzle-to-nozzle pressure drop:
67.2 psid e Initial core flow at 10,082 gpm o Primary hot leg temperature at 750*F Secondary loop flow at 75 +1% of 13,200 spm e
e Secondary cold leg temperature at 602*F All six heat transport system (HTS) pony motors de-energized (subsequent to e
analyses reported here, current FFTF plans call for one intemediate pony motor to be operating).
4 RESULTS OF CALCULATIONS Figures 2 through 5 present the calculation results in terms of coolant tempera-tures at the axial level where maxinum temperatures occur which corresponds to the top of the active fuel section of the Row 2 and Row 6 F0TA's (applicable to two T/C locations in the FTR, as shcun in Figure 1).
For each F0TA, there are six cases analyzed. The "1 Hr-NCM", "56 Hr-NOM" cases (defined in Section 3) represent the nominal case sodium temperatures.
The actual temperature in these channels would be expected to correspond with these two curves, depend-ing upon the irradiation history.
Also shown in Figure 2 are the results of the initial base run described in Section 2.
Due to uncertainties, as dis-cussed in Section 2, two additional curves are presented in Figures 2 through 5 representing the "1 Hr-UNC" and "56 Hr-UNC" cases defined in Section 3.
The actual temperatures at the specified locations should not exceed these two curves representin'g (3a) uncertainty cases. Thermocouple measurement uncer-tainties and time delays are not included in the predictions.
Thus, the acceptance criterion is that if the maximum measured temperature at specified
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channels (corrected for above instrument effects and actual boundary conditions) is less than the curve, it would follow that the DEM0/ COBRA-WC/ FORE-2M calcu-lations made with design data conservatively envelope the temperature response of the natural circulation test. Also shown in Figures 2 through 5 is the i
i IV-G
maximum temperature based on the current CRb3P design method (CRBRP/ PRE-NCVP).
This method uses DEMD calculated reactor flows based on maximum system AP's and maximum decay heat, does not consider the benefits of core inter-and intra-assembly flow and heat redistribution, and combines the hot channel factors conservatively.
5.
CONCLUSIONS The following conclusions can be made from the results obtained in this report:
a) The transients of the cases studied are very mild since the expected (nominal) peak transient AT's are less than 55 F higher than the initial steady state temperature rises as shown in Figures 3 and 5 for two different F0TA's and the maximum irradiation history, b) The accuracy of the predictions is expressed in terms of hot channel factors as conventionally used in reactor design.
By considering 3a uncertainties statistically, the maximum temperatures or '.nese two F0TA's are also shown in Figures 2 through 5.
The measured temperatures in the respective F0TA's should therefore not exceed "I Hr-UNC" or "56 Hr-UNC" curves with a 99.9% confidence level depending on the irradiation history of the core before shutdown.
The "56 Hr-UNC" case yields the highest temperature.
c) The CRBRP/ PRE-NCVP approach predictions are very conservative with respect to the nominal temperatures of the FFTF FOTA under natural circulation conditions.
IV-7
~~
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1100 NOTE: TEMPERATURES PREDICTED FOR T/C LOCATIONS A & B (FIGURE 1)
CASE "I HR - CRBRP/ PRE-NCVP" p
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Figure 3. Sodium Temperatures at Top of Actise Core Axial Position for Row 2 FOT.\\ with 56 Ilour Irradiation Ilistory (Test Initiated from 35'] Power /75'4 Flow) 5005 2 IV-9
1100 NOTE: TEMFFRATURES PREDICTED FOR T/C LOCATIONS C & D (FIGURE 1) 1000 CASE "1 HR - CRBRP/ PRE 41CVP" C
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l 1100 NOTE: TEMPERATURES PREDICTED FOR T/C LOCATIONS C & D (FIGURE 1)
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Figure 5. Sottium l'emperatures at Top of Actisc Core Axial Position for Row 6 FOTA willi 56 Ilom la.uli.iiion Ilistory iTest initiateil from 35'; l'owerj75'; Flow )
5005-4 IV-11
REFERENCES 1.
R.D. Coffield, et al., "LMFBR Natural Circulation Verification Program (NCVP) Review of Experimental Facilities and Testing Recommendations",
WARD-NC-3045-1, July 1977.
2.
R.D. Coffield, R.A. Markley and E.U. ' Khan, " Natural Convection Analyses and Verification for LMFBR Cores" International Working Group on Fast Reactors Specialists Meeting on Thermodynamics of FBR Fuel Assembly Under Nominal and Non-Nominal Operating Conditions, IWGFR/29, February 1978.
3.
R.M. Singer and J.L. Gillettee, " Measure'ments of Subassembly and Core Temperature Distributions in an LMFBR", AIChE Symposium Series, 73,
- p. 97 (1977).
f 4.
A.K. Agrawal, et al., " Dynamic Simulation of LMFBR Plant Under Natural Circulation", 18th National Heat Transfer Conference, ASME Paper 79-HT-6, August 1979.
5.
W.H. Allison, et al., "CRBRP; LMFBR Demo Plant Simulation Model (DEMD)",
CRBRP-ARD-0005, February 1978.
6.
T.L. George, et al., " COBRA-WC: A Version of COBRA for Single-Phase Multi-Assembly Thermal-Hydraulic Transient Analysis", PNL-3259, July 1980.
7.
J.V. Miller and R D. Coffield, " FORE-2M: A Modified Version of FORE-II Computer Prcgram for the Analysis of LMFBR Transients", CRBR,P-ARD-0142, November 1976.
8.
R.D. Coffield, et al., " Buoyancy Induced Flow and Heat Redistribution During LMFER Core Decay Heat Removal", Proceedings of Specialists Meeting on Decay Heat Removal and Natural Convection in FBR's, Brookhaven National Laboratory, NY, February 1980.
9.
H.P. Planchon, W.R. Laster and R. Calvo, "DEM0 Pre-Test Predictions for the FFTF Transient Natural Circulation Tests", WARD-94000-00321, March 1980.
10.
T.L. George, K.L. Basehore and W. A. Prather, " COBRA-WC Model and Predic-tions for a Fast Reactor Natural Circulation Transient", AIChE Symposium Series, 76,, No.199, p. 205 (1980).
11.
" User's Guide for Irradiation Experiments in the FTR", HEDL-MG-22, Rev. 2 Fby 1978.
12.
" Verification of Natural Circulation in Clinch River Breeder Reactor,"
transmitted to NRC via letter dated June 21, 1976.
IV-12
APPENDIX A DESCRIPTION OF FTR CORE CONFIGURATION AND SELECTED INPUT DATA 4
The schematic diagram of FTR is shown in Figure A-1.
The core map is shown in Figure A-2.
Approximatel;r 1/6 of the core is modeled in the COBRA-WC code, while FORE-2M uses an annular model of a single rod in the assembly (F0TA's).
Each F0TA is a 40-ft. long core assembly, consisting of a lower.12-ft. in-core fueled section and an upper 28-ft. instrument stalk section.
The lower section is essentially the same as a regular core driver except for the addition of pin bundle thermocouples in the F0TA.
The locations of these two F0TA's with-in the core are shown in Figure A-2.
The Row 2 F0TA is surrounded by 6 driver assemblies at two of its faces.
Of these two assemblies, inter-assembly heat
- transfer should be great,er for the Row 6 F0TA.
Table A-1 presents the 56-hour decay heat data obtained from HEDL.
Table A-2 indicates typical results from DEMD for the transient reactor inlet nozzle flow for the various conditions evaluated.
This information was used in performing the analyses for Figures 2, 3, 4 and 5 as described in Section 2 of this report.
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IV-15
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TABLE A-1 56-HOUR DECAY HEAT DATA *
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TIME (SEC.)
FUEL DECAY **
NON-FUEL DECAY **
0.0 0.05000 0.02147 1
1.0 0.05000 O.02147
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3.0 0.04514 0.01958 6.1 0.04032
( 0.01769 12.0 0.03692 O.01634-24.0
, 0.03315 0.01486 42.0 0.02994 0.01359 60.6 0.02784 0.01277 78.8 0.0263 0.01215 97.0 0.0251 0.01168 115.2 0.02412 0.01129 133.3 0.02332 0.01097 151.5 0.02263 0.01069 169.7 0.02204 0.01046 187.9 0.02153 0.01025 206.1 0.0210 0.01004 Delayed neutron power same as for 1-hour case from Reference 9.
Fr. action of steady state operating power.
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4 IV-16
s TABLE A-2 TYPICAL DEM0 OUTPUT DATA FOR FIGURES 2,3,4_AND 5 ANALYSES DEMDREACTORINLETN0ZZLEiFLOW,(lb/sec)
TIME (SEC.)
RUN 1 RUN 2 RUN 3 RUN 4 RUN 5
~
0.0 3668.0 3668.0 3668.0 3668.0 3668.0 10.0 1584.7 1584.6 1379.0 1584.6 1581.5 60.0 256.96 259.6 161.3' 260.4 266.2 100.0 50.6 55.1 44.0 59.2 58.2 155.0 55.7 62.1 54.3 68.2 64.8 200.0 65.2 72.7 63.0 79.2 75.3 o Run 1:
1-hour nominal conditions; e Run 2:
56-hour nominal conditin1s; e Run 3:
Same as Run 2, but with AP uncertainty; e Run 4:
Same as Run 2, but with decay heat uncertainty;
~
e Run 5:
Same as Run 2, but with recent FFTF shield / orifice l
assembly pressure loss coefficient data from HEDL.
NOTE:
Core inlet temperature can be maintained at 631*F for the first 200 seconds based on above DEM0 studies.
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NO. OF PAGES
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HARD COPY AT:
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ATTACHMENT VI MICROFICHE OF THE DEMO CODE (FFTF MODEL) -
PROGRAM MODIFICATIONS, INPUT AND OUTPUT DATA l
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gw ATTACHMENT VI MICROFICHE OF THE DEMO CODE (FFTF MODEL) - PROGRAM MODIFICATIONS MADE SINCE SUBMITTAL BY THE REFERENCE LETTER, EDITED INPUT AND OUTPUT DATA
Reference:
Letter S:L:890, P. S. Van Nort to R. P. Denise, "Trans-mittal of DEMO Computer Code, Revision 4," dated April 2, 1976.
Three sets of microfiche related to WARD-94000-00321 report are enclosed:
)
(1) 35% power, 75% flow best estimate case (2 pages)
(2) 35% power, 75% flow design case with 125% nominal decay heat (2 pages)
(3) 35% power, 75% flow design case with 75% nominal decay heat (2 pages) 4 e
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VI-1 l