ML19345B783
| ML19345B783 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/28/1980 |
| From: | ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | |
| Shared Package | |
| ML19345B780 | List: |
| References | |
| NUDOCS 8012020504 | |
| Download: ML19345B783 (5) | |
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ATTACHMENT I PREDICTION OF FAST FLUX TEST FACILITY NATURAL CIRCULATION TEST RESULTS USING CLINCH RIVER BREEDER REACTOR PLANT s PROJECT METHODOLOGY ,
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): ~9 21 PREDICTION OF FFTF HATURAL CIRCULATION TEST RESULTS USING CRBRP PROJECT
, ftETHODOLOGY The objectiva of the CRBRP Natural Circulation Verification Program (NCVP) is to analyticaily and experimentally verify numerical models for the des-cription of the thermal-hydraulic behavior of LMFBR plants when making transition to or oparating in the natural circulation mode. This verifica-tion is an important part of demonstrating the capability of adequate decay heat removal by means of natural circulation in LMFBR's. A description of the NCVP program was provided to NRC in a letter dated June 21, 1976.
In addition, the program approach to meeting this objective has been
. extensively published through various open literature publications and through conferences (Refs. 1, 2, 3, 4). A major component of the NCVP is comparison of analytical predictions with experimental data for experi- -
ments in facilities other than CRBRP. Planned natural circulation test-ing at the Fast Flux Test Facility (FFTF) presents an excellent opportunity to compare Project predictions with experimental data. A program of pre-test and post-test prediction of the FFTF tests is in progress. This attach-ment provides an overview of the analytical tools, approaches and data uti-lized for the FFTF test predictions. To the extent practical, the tools, approaches and data are the same as tnose to be used for prediction of natural circulation scenarios in CRBRP.
The corr ' uter codes utilized in the p-edictions are DEMO, FORE-ZM and COBRA-WC.
Tnese 3 codes, not a single code, were developed for performing the desired predictions because a single code would be too large to be of practical use for design applications. A brief description of each code and its role in the prediction follow ~s:
A. DEY0 (Ref. 5)
DEMD, a system-wide code, predicts the coupled performance of the reactor, heat transfer systems, and the steam generating system (including piping and plena heat capacity effects). To provide accurate results for the plant as a whole, yet naintain the code as an amenable tool with regard to computer storage requirements and running time, localized phenomena (which do not affect the
- system as a whole) are not given detailed modeling. An example of this is e item such as local flow / heat redistribution between and within core assem-
.,:ies at low flow. DEMD employs a three region model (fuel, non-fuel and bypass)
, for the core. To consider these localized phenomena with the required resolu-tion takes a separate computer code itself (i.e.. COBRA-WC was selected for this purpose). In general then, given a natural circulation event, DEMD pro-vides a verified prediction of the overall system state variables such as net flow through the reactor and bulk temperatures entering and exiting the core.
B. COBRA-WC (Ref. 6)
The CSSRA-WC code, which accounts for core inter- and intra-assembly flow and I heat redistribution, predicts the boundary conditions for a peak rod in the fuel and blanket assemblies given the reactor boundary conditions such as total reactcr inlet flow and bulk temperature from DEM0, individual core assembly powers, and individual core assembly thermal-hydraulic information.
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t C. FORE-2M (Ref. 7)
Given the locali ed peak channel flows and heat interchange as a function of tine, FORE-2M eredicts the hot channel coolant temperature for natural circula-tion which includes considering uncertainties in the predictions. This hot channel coolant temperature is used as the basis for determining the accepta-bility of natural ci-culation for decay heat removal of U1FBR's. To have a verified hot rod prediction, one must properly account for such effects as:
a) fuel restructuring; b) fuel-cladding gap conductance; c) stored heat in the rod; d) statistical significance of physics and engineering uncertainties in power, flow, dimensions, properties, etc.; e) localized hot spots on the rod due to the wire wrap or pellet eccentricity; f) localized decay heat variations; and g) localized rod power varia*. ions in the assembly.
Since FORE-2M is used for core hot rod analysis with input from DEMD and COBRA-WC, extreme detail in nodalization can be used to assure accurate tem-perature predictions. This level of detail cannot be part of the COBRA and DEMD predictions since the storage requirements and computer runnino times -
, would make the codes impractical. Thus, FORE-2M is t sed to predict verified hot channel coolant temperatures. [ Note: In addition, the code provides a detaileo nuclear physics prediction for the prompt and delayed neutron fission power as a function of time from shutdown with the capability of considering Doppler, sodium density, and radial / axial core expansion feedback plus the effects of control system worth and PPS functions. This information is used to verify the DEMD prediction of core power versus time].
The above analysis procedure is outlined by Figure 1.
. The planned FFTF tests encompast the transition into natural circulation cooling from a series of steady state initial conditions. All the tests will be in'+iated with a reactor scram and main coolant pump trip. The tests will include:
A) A transition to natural circulation in the primary loop from 5% reactor power, 75% primary loop flow. The secondary loop pumps will coast down to 10% speed at which time pony motors engage and provide 105 secondary
, loop forced flow.
B) A transition to natural circulation fri both primary and secondary loops from 35% reactor power and 75% flow.
C) A transition to natural circulation in both primary and secondary loops from 75% reactor power and flow following test subset B.
D) A transition to natural circriation in both primary and secondary loops from 100% reactor power and flow.
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The parts of this testing that are of most direct interest for fiCVP cualifica- ;
tion are the cransients run from the three normal power points, i.e.,100 ;,
s 75% and 35% reactor power - test subsets B, C and D above.
The pre-test predictions of the tests consists of performing nominal best-estimate calculations plus uncertainty evaluations of the nominal calculation.
The reactor flows and inlet temperature versus time for each of these tests is calculated with an FFTF version of the DEM0 code which substitutes a dump heat exchanger for the steam generator. The post-test analysis will assess the effects of actual test conditions as opposed to those used in the analyses, effects of measured uncertainties and consistency of the models in DEM0, COBRA-WC and FORE-2M with test results.
Two basic boundary conditions employed in the pre-test predictions are power history prior to scram (i.e., the decay powers) and the DHX sodium outlet temperature. When the actual values for these parameters are input to the programs used for the pre-test predictions, the results should agree closely with the actual measurements.
In addition, the post-test analysis will include an assessment of the accuracy of the data used in the pre-test predictions. The objective will be to deter-
'mine, to the extent possible with FFTF instrumentation, how much of the difference bet'.;een the best-estimate calculations and the measurements is due to uncertainties in the data used in the predictions.
REFERENCES
- 1. R.D. Coffield, et al ., "LMFBR Natural Circulation Verificaticn Program (NCVP) Review of Experimental Facilities and Testing Recommendations",
war 0-NC-3045-1, Jt,'y 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.D. Coffield, et al., " Buoyancy Induced Flow and Heat Redistribution During LMFBR Core Decay Heat Removal", Proceedings of Specialists Meeting on Decay Heat Removal and Natural Convection in FBR's, Brookhaven National Laboratory, NY, February 1980.
- 4. A. A. Bishop, R.D. Coffield and R. A. Markley, " Review of Pertinent Thermal-Hydraulic Data for LMFBR Core Natural Circulation Analyses", AIChE Sym-l posium Series, Vol . 76, pp.193-204,1980.
l 5. W.H. Allison, et al . , "CRBRP; LMFBR Demo Plant Simulation Model (DEM0)",
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.
i /. J.V. Miller and R.D. Coffield, ' FORE-2M: A Modified Version of FORE-II Computer Program for the Analysis of LMFBR Transients", CRBRP-ARD-0142, November 1976.
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