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BETHSY 6.2TC Test Calculation with TRACE using 3D Vessel Component

Andrej Prošek Jožef Stefan Institute

Jamova cesta 39 SI-1000, Ljubljana, Slovenia

andrej.prosek@ijs.si

Ovidiu-Adrian Berar adrian.berar@ijs.si

ABSTRACT

Recently, several advanced multidimensional computational tools for simulating reactor system behaviour during real and hypothetical transient scenarios were developed, including TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The purpose of the present study was to assess the 3D capability of the TRACE on BETHSY 6.2TC test. In addition, the TRACE 1D calculation and RELAP5 1D calculation were performed. The latest TRACE V5.0 Patch 03 and RELAP5/MOD3.3 Patch 04 thermal-hydraulic computer codes were used for calculations. BETHSY 6.2TC test is 15.24 cm equivalent diameter horizontal cold leg break on BETHSY facility. The BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. In general, all presented code calculations were in good agreement with the BETHSY 6.2TC test. The TRACE 1D calculation results are comparable to RELAP5 1D calculated results. With TRACE 3D calculation no significant improvement has been obtained. One reason for comparable results is already good agreement of RELAP5 and TRACE 1D calculations with experimental data, giving not much space to further improve the results. Nevertheless, it is important the finding that TRACE 3D calculation also reproduces the experimental data.

1 INTRODUCTION

Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. The TRAC/RELAP Advanced Computational Engine (TRACE) [1] is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package) [2]. It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc.

The TRACE code has 3D capability for vessel components, while U.S. NRC RELAP5 is one dimensional code. The developers stated that TRACE has superior capabilities and accuracy for most applications compared to RELAP5. The comparison between RELAP5 and TRACE code with 1D vessel model has already been done [3], [4]. The TRACE 1D calculation results for main safety parameters were as good as or better than the RELAP5 calculated results. The aim of this study is to assess the 3D capability of the TRACE on BETHSY 6.2TC test. The 3D reactor vessel was modeled manually using SNAP. The vessel component was then inserted into TRACE 1D input model. Namely, the TRACE input deck

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was previously semi-converted (using SNAP and manual corrections) from the legacy RELAP5 input deck, and was used as starting point [4]. This means that the geometry except the reactor vessel and renodalization done for TRACE is basically the same for RELAP5, TRACE 1D and TRACE 3D model, what gives very good basis for the comparison of the codes.

2 METHODS

In the following subsections the BETHSY facility and BETHSY 6.2TC test scenario are described first. Then the RELAP5 and TRACE input models are described, while for RELAP5, TRACE and SNAP code description the reader can refer to [1], [2] and [5], respectively.

2.1 BETHSY facility description

BETHSY was an integral test facility, which was designed to simulate most pressurized water reactor accidents of interest, study accident management procedures and validate the computer codes. The BETHSY facility was a scaled down model of three loop Framatome (now AREVA NC) nuclear power plant with the thermal power 2775 MW. Volume, mass flow and power were scaled to 1:96.9, while the elevations and the pressures of the primary and secondary system were preserved [6]. The core power has been limited to approximately 10% of nominal value, i.e. 3 MW. This means that the power was limited to the decay heat level and the transients without reactor trip could not be simulated. The design pressure on the primary side was 17.2 MPa and on the secondary side 8 MPa. Like in the reference reactor, the BETHSY facility had three identical loops, each equipped with a main coolant pump and an active steam generator. Every primary and secondary side engineered safety system was simulated. This included high and low pressure safety injection systems, accumulators (one per loop), pressurizer spray and relief circuits, auxiliary feedwater system and steam dumps to the atmosphere and to the condenser.

2.2 BETHSY 6.2TC test description

BETHSY 6.2TC test was a 15.24 cm (6 inch) cold leg break in the loop one without available high pressure and low pressure safety injection system [7]. Accumulators were available in the intact loops. The main aims of this test were to compare the counterpart test data from BETHSY and LSTF facilities and qualification of CATHARE 2 computer code. The experiment scenario was the following: opening of the valve simulating the break in the cold leg no. 1, accumulator injection in the intact loops when a primary circuit pressure was lower than 4.2 MPa and end of transient, when the primary circuit pressure was below 0.7 MPa.

2.3 Input model descriptions

The base RELAP5/MOD3.3 Patch 04 input model was used, which was continuously developed at Jožef Stefan Institute. The latest RELAP5 model consists of 398 volumes, 408 junctions and 402 heat structures.

The TRACE input model was converted from RELAP5/MOD3.3 input model using IJS procedure for RELAP5 to TRACE conversion using SNAP [9]. The SNAP conversion to TRACE mostly preserved the RELAP5 numbering of components (refer to [3]). The converted TRACE input model consisted of 172 hydraulic components and 72 heat structures. Finally, it should be noted that SNAP does not have the capability to convert one-dimensional

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reactor vessel model into three-dimensional reactor vessel model. This has to be done manually.

To get TRACE 3D input model the three dimensional pressure vessel was created using TRACE vessel component. In addition, the external downcomer was slightly modified. The adaptations were done manually using SNAP. The TRACE nodalization of BETHSY facility using 3D vessel is shown in Figure 1. The model consists of 149 hydraulic components and 75 heat structures. The number of hydraulic components (not the cells) in 3D model is decreased compared to 1D model because vessel represents one hydraulic component. The vessel component consists of 31 axial levels, 5 radial rings and 3 azimuthal sectors. The core region consists of 12 axial levels, 2 radial rings and 3 azimuthal sectors. Each azimuthal sector within each ring has its own heat structure representing heater rod. This gives in total six heat structures, which can be seen from Figure 1 (on the top of reactor pressure vessel).

Figure 1: Bethsy facility input model for TRACE using 3D vessel component

3 RESULTS

3.1 BETHSY 6.2TC steady-state calculations

Table 1 shows initial and boundary conditions for BETHSY 6.2TC test. The RELAP5 model was initialized to cold leg temperature; therefore the secondary pressure is not exactly matched. The difference comes from the geometry and the code models. The steam generator levels and masses were matched to average measured values. The pressurizer pressure and level were also matched to average measured values. The core power was input value. Also the TRACE 1D model was initialized to cold leg temperature. The TRACE 1D input model has slightly different initial values for core outlet temperature, core flow and secondary pressure. The TRACE 3D model, which was built on TRACE 1D model, was also initialized with the artificial controllers. By better matching the total flow also the core outlet temperature is closer to experimental value. In general, the agreement of initial and boundary conditions is good for all calculations.

Finally, Figure 2 shows initial conditions for TRACE 1D calculation.

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Figure 2: TRACE 1D calculated initial and boundary conditions

Table 1: Comparison of initial conditions for BETHSY 6.2TC test

Parameter Measured RELAP5 TRACE 1D TRACE 3D core thermal power (kW) 2863 ± 30 2864 2860 2863 pressurizer pressure (MPa) 15.38 ± 0.15 15.38 15.38 15.38 pressurizer level (m) 7.45 ± 0.2 7.45 7.45 7.45

total flow (kg/s) 16.81 (calculated from core power)

16.84 16.61 16.81

core inlet temperature (K) 557.2 ± 0.4 557.2 557.2 557.2 core outlet temperature (K) 588.2 ± 0.4 588.1 588.8 588.3 secondary side pressure - per SG (MPa)

6.84 ± 0.07 6.83 6.69 6.70

steam generator level - per SG (m) 11.1 ± 0.05 11.1 11.1 11.2 feedwater temperature (K) 523.2 ± 4 523.2 523.2 523.2 downcomer to upper head flow (kg/s) 0.047 0.047 0.047 0.0003

3.2 BETHSY 6.2TC transient calculations

The transient results are presented in Table 2 and Figures 3 through 8. Table 2 shows the main sequence of events. As can be seen the RELAP5 calculation using standard BETHSY input model is in a better agreement with the experiment in the initial phase than TRACE 1D calculation using converted model, while in the later part the TRACE 1D model was better than RELAP5. The timing for TRACE 3D calculation is comparable to TRACE 1D calculation. The time sequence of events mostly depends on the break flow. For RELAP5 original Ransom-Trapp break flow model the values of 0.85, 1.25 and 0.75 were used for subcooled, two phase and superheated discharge coefficients, respectively. For TRACE 1D break model the values of 0.8 and 0.9 were used for subcooled and two phase discharge coefficients, respectively. For TRACE 3D break model the values of 0.8 and 0.85 were used for subcooled and two phase discharge coefficient, respectively. The values of break

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discharge coefficients for TRACE calculations were selected after some sensitivity studies to match the integrated break flow as closely as possible.

Table 2: Main sequence of events for BETHSY 6.2TC test

Events Time (s) Measured RELAP5 TRACE 1D TRACE 3D

Break opening 0 0 0 0 Scram signal (13.1 MPa) 8 2 3 4 Safety injection signal (11.9 MPa) 12 8 9 9 First core uncovery 92 90 136 126 Loop seal clearing 134 155 173 219 Primary/secondary pressure reversal 172 175 203 185 Second core uncovery 334 280 253 246 Accumulator no. 2 and 3 injection start (4.2 MPa) 345 365 329 368 Accumulator isolation no. 2 (no. 3) (1.5 MPa) 948 (976) 925 (925) 801 (802) 682 (648) Pressurizer pressure < 0.7 MPa 2065 2230 2167 2071

Primary pressure is shown in Figure 3(left). In spite of slightly larger RELAP5 break flow than TRACE break flow the pressure drop is faster in case of TRACE calculation. Secondary pressure is shown in Figure 3(right). The experimental results indicate that atmospheric relief valves were open a few tens of seconds and that pressure was not strictly kept to 7.3 MPa like prescribed by procedure [7] and simulated by calculations. Later, the agreement between experiment and calculation is slightly better for RELAP5 than for TRACE calculations. It should be noted that in general after initial period the secondary side has small influence on the primary side and by this on the overall calculation.

Figure 4 shows core inlet (shown left) and core outlet temperature (shown right). RELAP5 is in slightly better agreement. Regarding core outlet temperature the measured value is mixture temperature, while in the case of calculations liquid temperature is shown, therefore heatup in the last phase of transient is not seen.

In Figure 5 are shown the break flow and the integrated break mass flow. It can be seen that the calculated break flows are quite well matched, in the range of 10% uncertainty. The integrated break flow better agree for the TRACE calculation.

Figure 6 show the heater rod surface temperatures in the middle (shown left) and at the top of the core (shown right). The first heatup is predicted by all calculations at the top of the core, while in the middle of the core TRACE 3D did not predict the heatup. The time of the second heatup is better predicted by TRACE calculations, while the heatup rate was better in the case of RELAP5 calculation. The heatup rate for TRACE 3D model is significantly larger than in the experiment.

Figure 7(left) shows the differential pressure in the core, while Figure 7(right) shows primary side mass calculated based on the mass injected by accumulators and mass discharged through the break. TRACE 1D and 3D calculations are similar. In spite of correct TRACE calculated mass discharged through the break the TRACE calculated primary mass is smaller than the experimental. Similar is in the case of RELAP5 calculation.

Finally, the accumulator behavior is shown in Figure 8 showing the accumulator pressure (shown left) and the integrated accumulator injected mass (shown right). The accumulator injected mass is very close to experimental value in the case of all calculations. The trend for RELAP5 is better than for TRACE calculations.

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Figure 3: Pressurizer pressure (left) and steam generator 1 pressure (right)

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Figure 5: Break mass flow (left) and integrated break mass flow (right)

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Figure 6: Heater rod surface temperature at the middle of the core (left) and at the top of the

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Figure 8: Accumulator no. 2 pressure (left) and integrated accumulators injected mass (right)

3.3 Discussion

The results show that in general the RELAP5 and TRACE 1D calculation are comparable, being TRACE 1D slightly better. One reason may be that in the facility the phenomena were mostly one dimensional. Introducing TRACE 3D vessel component for pressure vessel required manual work. Besides the hydrodinamic vessel component also the vessel heat structures form significant part of input model. The calculated heater rod temperatures were higher than in the TRACE 1D calculation and this should be investigated in the future after this first study. One difficulty with constructing the 3D vessel was, that by setting the flow area fraction in the vessel, the desired core bypass flow could not be achieved. Therefore the core bypass area fraction was set to zero, i.e. no bypass flow was modelled. According to differential pressure in the core there is slightly higher core uncovery in the case of TRACE 3D calculation comparing to TRACE 1D calculation, but not such to cause so fast heatup as calculated. Therefore, some further investigation is needed regarding 3D vessel component model.

4 CONCLUSIONS

Three calculations of the BETHSY 6.2TC test were performed using the latest RELAP5 and TRACE code versions. The overall results obtained with TRACE 1D input model were similar to the results obtained by RELAP5 1D input model. Both in the TRACE 1D and RELAP5 calculation first core uncovery until accumulators started to inject were predicted. After emptying accumulators the second core heatup was predicted by all calculations due to second core boil off. It was shown that TRACE 1D results obtained by converted input model from RELAP5 to TRACE are as good as the RELAP5 results obtained by the original RELAP5 input model. The results of TRACE 3D calculation were comparable to TARCE 1D

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calculation except the heater rod temperatures. The timing is correct. However, some more investigation is needed for 3D vessel model, which seems to cause the differences in heater rod temperatures.

ACKNOWLEDGMENTS

The authors acknowledge the financial support from the state budget by the Slovenian Research Agency program no. P2-0026 and financial support from Slovenian Nuclear Safety Administration and Krško Nuclear Power Plant by project No. POG-3473.

REFERENCES

[1] U. S. Nuclear Regulatory Commission, “TRACE V5.0 User Manual”, Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, Washington, DC.

[2] APT, “Symbolic Nuclear Analysis Package (SNAP)”, User's Manual. Report, Applied Programming Technology (APT), Inc., February 2011.

[3] A. Prošek, O.-A. Berar, “Advanced presentation of BETHSY 6.2TC Test results calculated by RELAP5 and TRACE” . Sci. Technol. Nucl. Install. 2012, pp. 812130-1-812130-15.

[4] A. Prošek, O.-A. Berar, “BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code”, Proc. 14th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), , Toronto, Canada, September 25-30, 2011 NURETH-14. [S. l.]: Canadian Nuclear Society, 2011, pp. 150.1-150.10.

[5] USNRC, “RELAP5/MOD3.3 code manual”, Patch 04, Vols. 1 to 8, Information Systems Laboratories, Inc., Rockville, Maryland, Idaho Falls, Idaho, prepared for USNRC, 2010.

[6] CEA, “BETHSY, General Description”, No. SETh/LES/90-97, CEA (Commissariat à l'énergie atomique et aux énergies alternatives), Grenoble, France, April 1990.

[7] CEA, “BETHSY – Test 6.2 TC, 6 inch cold leg break counterpart test, Test analysis report”, No. STR/LES/91-034, CEA (Commissariat à l'énergie atomique et aux énergies alternatives), Grenoble, France, October 1991.

[8] A. Prošek, O.-A. Berar, “Analysis of BETHSY 9.1b test with different RELAP5 code versions” Proc. 20th International Conference Nuclear Energy for New Europe 2011, Bovec, Slovenia, September 12-15, 2011, Nuclear Society of Slovenia, 2011, pp. 812.1-812.8.

[9] A. Prošek, O.-A. Berar, “IJS procedure for RELAP5 to TRACE input model conversion using SNAP”, Proc. 2012 International Congress on Advances in Nuclear Power Plants, Chicago, Illinois, USA, June 24-28, American Nuclear Society, 2012, pp. 2032-2041.