ML20136H716
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| Issue date: | 12/23/1985 |
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| CON-FIN-A-6328 EGG-M-19685, PRUR-851223, NUDOCS 8601100469 | |
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V.7-1 A COMPARISON OF THREE TRAC-BWR ASSESSMENT STUDIES WITH DATA FROM A ROSA-III SMALL BREAK TEST K.C. Wagner Idaho National Engineering Laboratory P.O. Box 1625 Idaho Falls, Idaho 83415 (208) 526-0809 ABSTRACT A comparitive analysis of three assessment studies of Run 912 from the ROSA-III small break LOCA test series has been performed.
Each study was performed subsequent to the development of an improved version of the TRAC-BWR safety analysis code. The ability of each computer code to calculate the pertinent thermal hydraulic phenomena in the experiment is presented. The run time statistics are also given.
I.
INTRODUCTION AcomparisonofthreeassesgmentstudiesoftheRigofSafetyAssessment (ROSA)-III, Run 912 is made.
In each study, the most advanced version of assessedgientReactorAnalysisCodeforBoilingWaterReactors(TRAC-BWR)was the Tran 3 using Version 12 of The earliest study was performed in 1982 4
the TRAC-BWR. The second version of the code assessed was TRAC-BD1/M001.
TRAC-801/M001 refined many models from Version 12 and provided additional modeling capability. The most recent assessmert study used the TRAC-BF0 version of the code.
TRAC-BF0a used two step rumerics for all one dimensional hydraulic components to reduce comuter run time.
This paper summarizes the results from each of the three assessment studies.
The ablilty of the codes to predict the pertinant thermal hydraulic phenomena and the computer runtime statistics are discussed. A brief description of the facility and test is included to give necessary background information.
Descriptions of the computer models and respective codes are also presented.
II.
FACILITY AND TEST DESCRIPTION ROSA-III was a well instrumented experimental facility in Japan Atomic Energy ResearchInstituteinJapanjesignedtosimulateawiderangeofpostulated accident scenarios for a BWR. The facility was 1/424 volumetrically scaled version of a BWR/6. There were four half length, electrically heated power bundles that simulated a BWR nuclear core. One bundle simulated a high power channel while the other three simulated average power channels.
Each bundle contained 62 heated rods and 2 water rods in an 8 x 8 array.
The ROSA-III facility has a prototypical BWR/6 Emergency Core Cooling System (ECCS). The ECCS ircludes a High Pressure Core Spray (HPCS), a Low Pressure Core Spray (LPCS), and a Low Pressure Coolant Injection (LPCI). The HPCS was assumed failed for Run 912.
The Main Steam Line (MSL) at the top of the pressure vessel served three functions.
First, it simulated the resistance of the steam turbines during steady state operations.
Secondly, it acted as a Safety Relief Valve (SRV) during transient situations to maintain the system pressure below 8.47 MPa.
a TRAC-BF0 is a pre-released version of TRAC-BF1, which is scheduled for release in late FY 1986.
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y,7 2 Finally, it acted as an Automatic Depressurization System (ADS) during the Loss-of-Coolant-Accident (LOCA) to reduce the system pressure.
There were two recirculation loops attached to the ROSA-III pressure vessel.
Each loop had two jet pumps'and one recirculation pump. A prototypical BWR has the jet pumps inside the pressure vessel downcomer annulus. However, the ROSA-III jet pumps were placed external to the vessel due to modeling constraints. The break Iccation for Run 912 was in the recirculation loop and just upstream of the recirculation pump.
l Run 912 simulated a 5% break of the recirculation pump suction piping.
The transient was initiated when the break valve was opened.
The pump power
-was turned off at the. initiation of the transient and the core power was controlled to simulate a prototypical BWR power decay.
The system pressure decreased until the Main Steam Isolation Valve (MSIV) was closed at 24 s.
Following the MSIV closure, the system pressure rapidly increased until the SRV was manually operated to keep the system pressure below 8.47 MPa. The ADS valve opened at 158 s and caused a rapid depressurization.
The top of the heater rods dried out prior to ADS actuation.
The rods were wetted by lower plenum flashing upon the initiation of the ADS. From ' 5 s to 261 s, the entire core uncovered and more heat-ups were measured. The LPCS and LPCI were initiated at 318 s and 406 s, respectively. All the heater rods were quenched between 328 s and 433 s.
The peak cladding temperature occurred at 410 s at the midolane of the high power rod.
III. COMPUTER CODE AND MODEL. DESCRIPTIONS A chronology of the development of the three codes and models is described in this section. Each major version of the TRAC-BWR code was independently assessed using ROSA-III data. Consequently, a comparison of the assessment studies illustrates the development of the code to predict pertinent chenomena during a small break LOCA. A brief descriotions of the three versions of the computer code are presented followed by a discussion of the models used in the three calculations.
TRAC-BD1/{ersion12wasdevelopedfromaninterimversionoftheTRAC-PF1PWR LOCA code. This version was extensively modified to permit accurate modeling of a BWR. TRAC-BD1/ Version 12 was developed primarily as a large break LOCA code.
l TRAC-BD1/M001, released in April 1984, allowed a greater amount of flexibility in the types of accidents which could be simulated. While i
TRAC-BD1/ Version 12 was primarily a large break LOCA code, TRAC-BD1/M001 was l
improved to allow better simulation of small break LOCA's and operational transients. The major improvements include: a comprehensive control system, a two phase level tracking model, improved constitutive relationships between the fluid phases and the structure surface, and balance of plant modeling capability.
l The most recent version of the code, TRAC-GF0, uses a two-step numerical scheme.
The two step method allows the calculational time step to exceed the
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material Courant limit'in the one-dimensional components. A typical TRAC-BWR model uses one dimensional components in all areas except the vessel structure. Consequently areas in BWR models which limited the code to a small calculational time step in previous versions of the code did not
V.7-3 influence the time step in TRAC-BFO.
This permitted improved running time without compromising the integrity of the calculation.
The TRAC-801/ Version 12 vessel nodalization is shown in Figure 1.
The two dimensional vessel model featured twelve axial levels and three radial rings. No azimuthal dependency was modeled in the vessel. The two inner rings between levels four and seven represented the high and average power channel and bypass regions. The annular downcomer was modeled in the outer ring. The guide tubes and the channels were modeled internally in the vessel by using one dimensional components and source connections to the vessel.
-The addition of the level tracking model in TRAC-801/M001 permitted a coarser nodalization to be used in the vessel region. A fine nodalization was used in the TRAC-B01/ Version 12 model to prevent numerical diffusion of vapor and allow accurate modeling of the recirculation line connections to the vessel.
The level tracking model eliminates numerical diffusion and accurately models the source connection elevation inside of a vessel cell.
In the TRAC-801/-
1 M001 and TRAC-BF0 models, several levels were comoined to reduce the number of axial' levels from-twelve to nine.
The core regions in the TRAC-BWR models were simulated using a CHAN component. A CHAN component in ring 1 of the vessel simulated the high powered bundle, while another CHAN component in ring 2 modeled three average power bundles.
In the TRAC-BD1/ Version 12 model, four rod groups were used in the high powered CHAN component. Three groups corresponded to the three radial peaking factors found in the core while the fourth group simulated two water rods. After examining the results from the Version 12 calculation, it was determined that radiation was not the dominate heat transfer mechanism for the present test.
Therefore, only one rod group per channel was used in the TRAC-BD1/M001 and TRAC-BF0 models.
Both the intact and broken loops were modeled. One JET PUMP component was used in each loop to model two jet pumps in the facility. The 5% break plane was modeled using a TEE component.
Flow versus pressure tables from the test data were input as boundary conditions to the models. The LPCI and LPCS were activated at system pressures of 2.38 and 1.81 MPa, respectively.
Ambient heat loss from the facility was also modeled. No experimental data was available for the ROSA-III heat loss distribution. However, the total heat loss was estimated to be 150 kW. 'An error in the TRAC-B01/ Version 12 and TRAC-B01/M001 models over-estimated the amount of heat loss in the.
recirculation loops by a factor of two. The TRAC-BD1/ Version 12 model had a 100 kW loss from each loop and a 50 kW loss in the vessel.
The TRAC-BD1/M001 model used a 75 kW loss from each loop and a 75 kW loss from the vessel.
The percentage of heat loss at the vessel.was increased after the TRAC-B01/-
Version 12 assessment in an effort to improve the agreement with the system pressure. The TRAC-BF0 model used a 75 kW heat loss from the vessel and a 38 kW heat loss from each recirculation loop.
The sensitivity of the system repsonse upon the ambient heat loss is discussed in the Results Section.
IV.
RESULTS The results from the three assessment studies are presented in this section.
An uncertainty analysis was not performed on the data, consequently, the results are presented without uncertainty bands. However, an uncertainty l
l
t V.7-4 estimate was received for the break flow and is indicated on the figure.
Althougn no uncertainties were given for the heater rod temperature data, the data were processed by elevation and linear heat rate (high or low power channel) to determine the minimum and maximum data.
Both the curves are presented with the calculated results to illustrate the range of data.
i Several comparisons are presented to assess the ability of the codes to predict the pertinent thermal hydraulic phenomena during the experiment. The system pressure and heater rod responses were determined to be important i
parameters that characterize a small break simulation.
These parameters are directly influenced by an accurate calculation of the steam line flow, the break flow, the regional mass distribution, and ambient heat loss. To facilitate discussion of the comparisons, the results from the three studies will be referred to as V12 for the TRAC-B01/ Version 12 analysis, M001 for the TRAC-801/M001 analysis, and BF0 for the TRAC-BF0 analysis.
The calculated and measured system pressure response is presented in Figure 2.
The pressure remained fairly constant at the initiation of the transient until the core power decay began at 8.8 s.
Subsequently, the pressure decreased until the MSIV closure at 24 s and then rapidly pressurized to the SRV set point (8.40 MPa).
The experimental and V12 r
results showed a slight depressurization after 100 s until the ADS was activated at 158 s (164 s in V12).
The M001 and BF0 calculated pressures remained near the SRV set point until ADS actuation.
The break flow in the i
M001 and BF0 simulations was lower than the measured flow and the break flow calculated by V12 which appeared to contribute to the calculated pressure hold up.
1 After ADS actuation, there was a rapid depressurization until 318 s when the feedwater line flashed and the LPCS liquid fell into the core and vaporized.
This reduced the measured depressurization rate and delayed LPCI initiation until 406 s.
In all the calculations, the LPCS and LPCI activations were based upon system pressure trips (as in the experiment). Consequently, variations in the calculated depressurization rate from the measured rate caused variations in the initiation of the LPCS and LPCI safety systems.
The 1
V12 pressure calculation was in good agreem6nt with the data until 318 s.
At approximately 325 s during the V12 transient, the feedwater line began to i
flash and the LPCS was initiated. There was a slight change in the calculated depressurization rate at this time.
Closer examination of the V12 results revealed that the LPCS liquid injected into the upper plenum was held above the upper tieplate. Conversely, in the M001 and BF0 simulations, LPCS fluid drained into the core and was vaporized. This provided better agreement with the measured delay time between LPCS and LPCI initiation by slowing the depressurization rate. However, all versions of the code calculated smaller delay times than were recorded during the experiment.
The M001 depressurization rate was in good agreement with the data. Due to the pressure hold up prior to ADS actuation, the blowdown timing was slightly shifted to the left of the data. As mentioned before, the calculated change and duration in the sytem depressurization after the feedwater flashed and LPCS initiation was slightly smaller than measured. A sensitivity study during the V12 assessment confirmed the importance of feedwater flashing to the change in depressurization rate after 318 s.
It appears that the feedwater line temperature in the calculations may have been slightly too w,-..,
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In fact, a 5 K decrease in temperature would delay flashing until the system pressure fell an additional 0.22 MPa.
However, no experimental data were available to check this hypothesis.
.The BF0 calculated depressurization rate was greater than the measured rate after approximately 200 s.
Examination of the calculation showed that system flashing in the loops supplied vapor to the ADS valve at the top of the vessel and reduced the core flow. The calculated reduction in the core flow caused the top of the core to dry out and heatup. The premature DNB at the top of the core reduced the steam production and increased the calculated depressurization rate. This was not observed in the other calculations or the experiment and was attributed to lower heat loss in the recirculation loops during the BF0 simulation. A comparison of the M001 and BF0 calculations showed that higher heat loss in the M001 calculation delayed system flashing by 30 s and provided better agreement with measured depressurization rate.
Since there is a high degree of uncertainty in modeling the ambient heat loss distribution, it is suspected that the BF0 loop heat loss was too low.
i The break mass flow is presented in Figure 3.
The measured break flow had a long subcooled blow down until the ADS system was initiated.
Subsequently, the flow transitioned to saturated flow for the duration of the experiment.
The M001 and BF0 results also show transition to saturated flow upon ADS actuation. However, the V12 break flow became saturated 47 s prior to ADS l
activation. The TRAC-B01/ Version 12 subcooled critical break flow model oroduced unstable results during this small break simulation. Consequently, a characteristic analysis of the break flow was used until ill s when the flow became saturated. The subcooled model was revised subsequent to the V12 simulation and worked satisfactorily in the M001 and BF0 simulations.
Based upon an examination of the calculated and measured system pressure responses, it appears that the V12 subcooled break flow was slightly too high and the M001 and BF0 subcooled break flow was too low. However, due to the large uncertainty in the break flow data, no definitive conclusions could be reached.
A comparison of the high power channel thermocouple data at the core midplane with the calculated rod surface temperatures is shown in Figure 4.
The measured results remained near saturation conditions until 220 s and then departed from nucleate boiling conditions. The maximum heater rod data show a reduction in the heat-up rate after LPCS initiation (335 s) and a rapid quench shortly after 406 s when the LPCI fluid flooded the core.
The M001 and BF0 calculated temperature responses were in good agreement with the measured trends whereas the V12 results show two additional heat-ups. A discussion of each calculation's heater rod response follows.
The V12 results had three heat-ups during the transient.
The first heat-up was caused by an under prediction of the core inlet flow during the pump coast down. The ROSA-III external jet pumps were atypical of a prototypical BWR and were not satisfactorily modeled in the V12 code. Momentum from the suction source was not considered in the solution of the jet mixing equation. Consequently, the core inlet flow was too low and a core heat-up was calculated.
The second heat-up was caused by restrictive counter-current flow limiting at the core upper tieplate. No liquid was nllowed to drain below the upper tieplate from the upper plenum and the second rod heat-up was
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- V.7-6 calculated.
In the other calculations and the measured data, the rods in the upper portion of the core showed heat-ups. However, this did not occur until all the liquid above the tieplate had drained into the core. Upon ADS initiation (165 s), the heater rods were quenched prior to the final heat-up. The V12 results show a reduced heating rate which was attributed to an over prediction of the convective heat transfer in the core. The calculated peak rod cladding temperature was 665 K.
The core was quenched in
'the V12 calculation upon LPCI initiation.
The M001 heater response was in excellent agreement with the data. The MOD 1 heater rod temperatures and: core inventory calculations represent a significant improvement over the V12 simulation. The improvements in the interfacial drag package, the jet pump models, and the heat transfer relation-stips, provided a better simulation of phenomena. A comparison of the MOD 1 and measured core liquid levels revealed an exec 11ent agreement between the M001 and the measured results. The maximum cladding temperature of 748 K at 394 s was in excellent agreement with the range of data measured at that location. The MODI core was calculated to quench shortly after LPCI initiation.
The 8F0 results were also in good agreement with the measured data. However, due to the reduction in core flow after 195 s, the final heat-up was slightly premature.
In general, the calculated trends were in excellent agreement with the M001 and measured results. The peak cladding temperature of 755 K was in excellent agreement with the range of experimental data.
V.
CODE RUN TIME STATISTICS The code'run time statistics will be discussed briefly in this section. An indication of how fast a computer code runs is found by taking the ratio of the computer central processor unit (CPU) to transient real time (RT) ratio.
The CPU /RT ratios ranged from 65 for V12 to 18 for 8FO. The cost was greatly reduced in each successive calculation.
The major improvements from the.V12 to the M001 calculation was attributed to a coarser noding in the vessel due to the addition of the level tracking model. The improvements from the M001
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to 8F0 calculation were attributed to the fast numeric. scheme incorporated in TRAC-8FO. Although the BF0 calculation ran much faster, several problems prohibited the calculation from running even faster. These problems were identified to the TRAC-8WR code development group and are being addressed.
The CPU /RT values are summarized in Table V.7-1.
VI.
CONCLUSIONS The overall trends of the Run 912 transient were well predicted by the three respective codes. Each code predicted the pertinent thermal hydraulic phenomena and the sequence of events. Many deficient areas were identified l
in the V12 simulation that were corrected in the subsequent versions. Close examination of the three assessment studies revealed several modeling sensi-l tivities which effect the calculation of small break transient. A significant improvement in run time was observed with each successive version i
of the code.
I VII. REFERENCES 1.
Yoshinari Anoda et al., Experiment Data of ROSA-III Integral Test Run l
l 912 (5% Split Break Withcut HPCS Actuation), JAERI-M 82-010, 1982.
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i V.7-7 2.
INEL TRAC-BWR Development Group, TRAC-801/M001: An Advanced Best Estimate Computer Code.for Boiling Water Reactor Analysis, NUREG/CR-3633, April 1984.
3.
R. J. Dallman, TRAC-801 Calculation and Data Comparison of International Standard Problem 12, EGG-CAAD-5860, May 1982.
4.
K. C. Wagner, TRAC-801/M001 Assessment Using ROSA-III Small Break Data, EG&G Report, Fin A6328, September 1984.
5.
K. C. Wagner, An Assessment of TRAC-8F0 Using ROSA-If! Data, EGG-RST-6881, June 1985.
Table V.7-1.
SUfHARY OF RUN TIME STATISTICS Parameter TRAC-801/V12 TRAC-801/M001 TRAC-8F0 CPU /RT 65 35 18 Number cells 142 131 131 liL !UL j
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NOTICE This paper was prepared as an account of work soonsered by an agency of the United States Government. Neither the United States Government nor any agency therof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this paper, or represents that its use by such third party would not infringe privately owned tights. The views expressed in this paper are not necessarily those of the U.S. Nuclear.
Regulatory Commission.
Work supported by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research under DOE Contract No. DE-AC07-76I001570.
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